Safety Analysis Code Research Articles

A three-dimensional criticality safety analysis code for spent fuel solution system

The nuclear criticality safety analysis is crucial for the nuclear safety of spent fuel reprocessing plants. Thus, a set of numerical tools with high accuracy and efficiency to predict the criticality and simulate the hypothetical accidents in the reprocessing procedure is of great importance. Among them, this paper focus on the analysis of the storage tank which is used for dissolution and storage of the spent fuel. According to the characteristics of spent fuel solution system, a parallel 3D critical safety analysis tool for fissile solution system, Hydra-TD, is developed. Generally, it contains the cross-section generation model, the three-dimensional space-time neutron kinetics model, and the R-Z two-dimensional heat conduction and radiolysis gas simulation model. The verifications based on the experiments of the SILENE and TRACY facility are conducted. The results show good accuracy of the prediction of the first fission power peak, multiplication time and total fission energy, indicating the reliability of these numerical models.

Validation and parametric study of FFRD model in DRACCAR code

The recent revision of Emergency Core Cooling System (ECCS) regulations in Korea has necessitated the consideration of Fuel Fragmentation, Relocation, and Dispersal (FFRD) phenomena in nuclear reactor safety analyses. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed and integrated the FFRD model into the domestically licensed safety analysis code, SPACE. Globally, US NRC’s FRAPTRAN and IRSN’s DRACCAR are available for evaluating FFRD phenomena. The FRAPTRAN code incorporates the FFR model, developed by Quantum Technology, while the fuel dispersal model is currently not included. In contrast, the DRACCAR code functions as an integrated analysis platform capable of modeling multi-dimensional thermal, hydraulic, mechanical, and chemical phenomena during a Loss of Coolant Accident (LOCA). This study conducts a thorough examination of the FFRD model in the DRACCAR code and validates its applicability through analysis using the Halden IFA-650 tests. The results demonstrate satisfactory predictive capabilities. Furthermore, a parametric study of key FFRD model parameters enhances the understanding of the FFRD model in the DRACCAR code. The development of detailed physical models in the FFRD model could significantly enhance the performance of the DRACCAR code, warranting the establishment of a comprehensive framework for these advancements. In the future, code-to-code comparisons between the DRACCAR code and other domestically developed integrated analysis platforms will be conducted to investigate various phenomena in depth.

Development of a phenomena identification and ranking table (PIRT) for intermediate break loss-of-coolant accident in PWRs

Recent safety issues such as cladding oxidation and fuel fragmentation, relocation and dispersal (FFRD) make the loss-of-coolant accident (LOCA) acceptance criteria more difficult to be satisfied. To obtain the adequate safety margin and more economical use of the nuclear power plant, re-classification of LOCAs for Korean operating PWRs is now under consideration to exclude the large break LOCA (LBLOCA) from the design basis accidents (DBAs). Therefore, the intermediate break LOCA (IBLOCA) might become the limiting break size of concern in the LOCA analysis. To accomplish this reform of LOCA classification, an extensive understanding of IBLOCA is crucial, and the applicability of safety analysis code should be confirmed. For this purpose, IBLOCA PIRT was developed. The PIRT panel was organized and the general process of the PIRT development established by Wilson and Boyack (1998) was adopted to develop the IBLOCA PIRT. Based on IBLOCA analyses using the safety and performance analyzing code (SPACE), the PIRT panel defined the temporal phases, systems, components and possible phenomena during an IBLOCA. For all possible phenomena, the relative importance and the knowledge level were determined for each component and each phase via discussions among PIRT panel members. For phenomena having relatively high importance, a code validation matrix was also developed. The results of the IBLOCA PIRT will be used to improve the SPACE code for IBLOCA application and resolve future regulatory issues.

Coupling of the best-estimate system code and containment analysis code and its application to TMLB’ accident

With the development of advanced pressurized water reactor technology, the thermal-hydraulic coupling effect between the containment and the primary system becomes increasingly tight. In order to meet the demand for integrated safety analysis between the containment and the primary system, this paper investigates a direct coupling method between the best-estimate system code Advanced Reactor Safety Analysis Code and the containment analysis program ATHROC (Analysis of Thermal Hydraulic Response Of Containment). The feasibility of this direct coupling method and the applicability of the coupled program for overall safety analysis are demonstrated using Marviken two-phase flow release experiments. The ATHROC/ARSAC coupled program is employed to analyze the impact of the pressure relief function of the CPR1000 nuclear power plant pressurizer on the behavior of the primary system and containment during the TMLB’ accident. The calculation results indicate that these measures can reduce the pressure of the primary system to the level acceptable by the low-pressure injection system, but at the same time, they cause the pressure in the containment to rise to nearly 0.4 MPa. Therefore, to ensure the structural integrity of the containment, it is necessary for the non-passive hydrogen recombiner to effectively reduce the hydrogen concentration, thereby avoiding additional pressure increase in the containment due to hydrogen deflagration, which could lead to overpressure failure. The findings of this study are of significant reference value for improving the safety performance of thermal-hydraulic systems in operational Gen-II and advanced Gen-III pressurized water reactor nuclear power plants.

Safety analysis code development and validation for lead-cooled fast reactor with helical coiled once-through steam generator

LFRs are one of six advanced Generation IV reactor concepts using lead-based coolant with high thermal conductivity, and HCOTSGs are generally adopted to highlight LFR miniaturization characteristics. An inhouse STH code SACLER with the HCOTSG model and the cover gas model was developed specifically for LFRs. NACIE facility was used to demonstrate SACLER accuracy. The HCOTSG model was verified using the RELAP code, experimental data, and ELSY design parameters. The validation results are consistent with the experimental data of NACIE. The HCOTSG model calculation results are in good agreement with the experimental data and the RELAP code. The error between the ELSY simulation result and the design value is within 1.0%. Liquid level simulation of ELSY shows that liquid levels are mainly determined by the pump head and flow resistance. SACLER can accurately simulate the LFR system with HCOTSG, including transient and steady-state conditions.

Verification of the SAM Code by Means of the Method of Manufactured Solutions: Application to Molten Salt Reactor Multiphysics Simulation

Software verification and validation constitute crucial phases in the development of simulation computer codes, particularly in the context of nuclear reactor safety analysis codes, where stringent safety requirements govern the development and deployment of nuclear technologies. This paper focuses on numerical verification study of the System Analysis Module (SAM) computer code, currently under development at Argonne National Laboratory. Specifically, we employed the Method of Manufactured Solutions (MMS) and proposed a verification technique tailored to the multiphysics simulation of molten salt reactors (MSRs). This research accomplished three main objectives. First, we have addressed key challenges associated with applying the MMS to MSR systems, arising from (1) the complex multiphysics coupling inherent in this problem and (2) the necessity to model the entire coolant loop for describing the drift of delayed neutron precursors outside the reactor core. The paper provides recommendations and guidelines to overcome these challenges, enabling the successful application of the MMS for simulating MSRs. Second, we have presented a comprehensive set of verification examples, serving as an exhaustive benchmark for code verification within the nuclear community. Third, we have established a robust verification of the SAM code’s capability to model MSR systems.

A methodology to quantify effects of constitutive equations on safety analysis using integral effect test data

To improve the predictive capability of a nuclear thermal hydraulic safety analysis code by developing a better constitutive equation for individual phenomenon has been the general research direction until now. This paper proposes a new method to directly use complex experimental data obtained from integral effect test (IET) to improve constitutive models holistically and simultaneously. The method relies on the sensitivity of a simulation result of IET data to the multiple constitutive equations utilized during the simulation, and the sensitivity of individual model determines the direction of modification for the constitutive model. To develop a robust and generalized method, a clustering algorithm using an artificial neural network, sample space size determination using non-parametric statistics, and sampling method of Latin hypercube sampling are used in a combined manner. The value of the proposed methodology is demonstrated by applying the method to the ATLAS DSP-05 IET experiment. A sensitivity of each observation parameter to the constitutive models is analyzed. The new methodology suggested in the study can be used to improve the code prediction results of complex IET data by identifying the direction for constitutive equations to be modified.

Model integration of the ex-vessel modules for the SFR safety analysis code SPECTRA

Model integration of the ex-vessel modules for the SFR safety analysis code SPECTRA

Machine Learning Applications and Uncertainty Quantification Analysis for Reflood Tests

The reflooding phase, a crucial recovery process after a loss of coolant accident (LOCA) in reactors, involves cooling overheated fuel rods with subcooled water. Its complex nature, notably in its flow regime and heat transfer, makes prediction challenging, resulting in high uncertainty and computation cost. In this study, we utilized the data assimilation (DA) technique to enhance the prediction of reflooding phenomena and subsequently deployed machine learning models to predict the accuracy of the safety and performance analysis code (SPACE) simulation. To generate the dataset for the machine learning model, we employed the sampling method for highly nonlinear system uncertainty analysis (STARU), providing a high-quality dataset for a complex problem such as a reflooding simulation. In this dataset, the physical models were assimilated under their selected uncertainty bands and utilized the effective sampling approach of STARU, generating the high-quality output and efficient enhancement of SPACE predictions. Consequently, the implemented machine learning model can be used to enhance model development and uncertainty quantification (UQ) analysis using the system code.

Proposal of Application of a Simple Analysis Code DYMOS for Accident Analyses of Molten Salt Reactors

The Molten Salt Reactor (MSR) possesses unique characteristics, including the circulation of liquid fuel salt in-core and ex-core, and the absence of cladding tubes of fuel rods. To perform safety analyses for accidents and transients, the selection of appropriate models is crucial. Proposed models vary from the coupling of a three-dimensional (3-D) nuclear reactor model and a detailed thermal/hydraulic (T/H) model to a simple point reactor model coupled with a lumped parameter model of the T/H system. The authors have been investigating the development of a simple and accurate model that can be utilized during the design stage and licensing evaluation, while also providing transparency, which means that models can be easily understandable by experts and reproduced in licensing evaluation. Given the distinctive features of the MSR, the peak heat flux or fuel cladding peak temperature is not a requirement. Instead, the most decisive parameter for safety evaluations is the fuel salt temperature in the fuel salt boundary. As demonstrated in this paper, the outlet of the MSR core consistently displays the highest temperature.Based on the afore-mentioned prospects for both nuclear and T/H systems, the authors have developed a simple safety and transient analysis code for MSR (DYMOS). The DYMOS code has been verified for MSRE experiments, as described in this paper. However, the previous verification was limited to small experimental MSRs, and there is a lack of verification for large reactor systems. This paper shows that the DYMOS code is applicable to these larger reactors. In other words, the main objective of this article is not to claim the originality of the model of DYMOS code, but to propose applicability of such simple code to large reactor systems.

A revised analysis towards accurate estimation of isothermal temperature coefficients in fast reactors having different power levels and core sizes

The accurate prediction of safety coefficient such as isothermal temperature coefficient of reactivity is an important requisite in the safe operation of any reactor and also in optimizing the reactor theoretically. At IGCAR, a multi-purpose safety analysis code PREDIS is being used for this purpose by employing the spatial distribution of first order perturbation worth as input. Estimated isothermal temperature coefficient is validated against the measured value of a small 40 MWt carbide core reactor FBTR and 400 MWt FFTF. Though the results are conservative, PREDIS is found to be under predicting the safety coefficients. A detailed parametric study showed that isothermal temperature coefficient is under predicted through PREDIS analysis because of neglecting the contributions of removal worth from the non-fuelled regions surrounding the core. Relative contributions of non-fuelled regions to isothermal temperature coefficient have been systematically quantified in different fast reactor cores such as 40 MWt carbide core of FBTR, 400 MWt FFTF, 1250 MWt oxide core of PFBR and 1500 MWt oxide core of FBR 1&2. From the study, the leakage contribution to the surrounding non-fuelled region is found to be significant in a smaller core, hence the removal worth of these regions should not be ignored while estimating the safety coefficients. Conservative assumptions of considering only the core region for the prediction of safety coefficient is assumed to be good, only for the medium sized reactors, where the leakage contribution to the isothermal temperature coefficient is not that prominent.

High-fidelity numerical investigation on structural integrity of SFR fuel cladding during design basis events

A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.

Preliminary development and verification of an OpenFOAM-based multi-physics coupling safety analysis code for Lead-based reactor

The thermal-hydraulics and safety analysis of Lead-based Fast Reactor (LFR) have drawn great attentions in recent years. By employing an implicit coupling strategy, a new multi-physics coupling code for safety analysis of LFR, was developed based on OpenFOAM. Point kinetics and fuel pin heat conduction modules were coupled to CFD module with turbulence consideration. Verifications were conducted by steady state and Beam Trip (BT), Beam Over Power (BOP) and Transient Over Power (TOP) accidents transient simulations in a single fuel pin channel of LFR. Results simulated by new coupling code proposed in this paper agree well with the ones provided by a safety analysis platform HC-PK-CFD based on ANSYS FLUENT. It demonstrated the feasibility and accuracy of implicit coupling strategy developed in this paper, as well as the capability of new program to perform multi-physics coupling simulations involved in thermal-hydraulics and safety analysis of LFR.

Analysis of dynamic responses of circulating fuel reactors using in-house 3D space-time kinetics code ARCH

Molten Salt Reactors (MSRs) are characterized by power generation in circulating liquid fuel salt rather than in static fuel pins of traditional nuclear reactors like in water cooled reactors. The fuel flow in MSRs drift the delayed neutron precursors (DNP) and results in lower delayed neutron fraction. Thus, MSRs show added complexity in dynamic analysis due to inherently coupled neutronics and thermal hydraulics. With regard to the reactor safety analysis, accurate predictions of DNPs distribution and its effect on reactor transients turn out to be an interesting subject of research. Due to these unusual features of MSRs, the traditional safety analysis codes are not applicable in dynamic analysis of reactors with circulating fuel. Challenges in accurate modelling and analysis of dynamic responses of MSRs have been described in present work using new capabilities in 3D space-time kinetics code ARCH. The numerical results of the implemented methodology presented in this study are compared with data available from Molten Salt Reactor Experiment (MSRE), a reactor operated at ORNL.

High-Fidelity Modeling and Experiments to Inform Safety Analysis Codes for Heat Pipe Microreactors

Heat pipe microreactors are reactor designs that primarily use liquid-metal heat pipes to cool the core. The main interest in heat pipes is the fact that they can remove heat passively. This, along with the use of liquid metal, allows the reactor to operate at higher temperatures. Although the use of heat pipes in nuclear reactors is new, liquid-metal heat pipe technology is mature. Nevertheless, experimental data on heat pipes are scarce, and very little is known about their behavior during abnormal operations and close to their thermal limits. Therefore, new experiments and accurate heat pipe simulations are needed to develop reliable closure models. This work describes a joint experimental and numerical investigation into heat pipes that attempts an initial closure of this gap. The numerical and experimental efforts are currently proceeding in parallel, aimed at different aspects of heat pipes. The numerical part is focused on gaps in local closures, and the experiments capture the overall heat pipe behavior.

Machine Learning Metamodel of a Computationally Intense LOCA Code

Abstract The nuclear power industry is increasingly identifying applications of machine learning to reduce design, engineering, manufacturing, and operational costs. In some cases, applications have been deployed and are providing value, particularly in data rich manufacturing areas. In this paper, we use machine learning to develop metamodel approximations of a computationally intense safety analysis code used to simulate a postulated loss-of-coolant accident (LOCA). Accurate metamodels run at a fraction of the computational cost (milliseconds) compared to the LOCA analysis code. Metamodels can therefore support applications requiring a high volume of runs such as optimization, uncertainty analysis, and probabilistic decision analysis, which would otherwise not be possible using the computationally intense code. In this study, training data is first generated by running the safety analysis code over a design of experiment. Exploratory data analysis is then performed followed by an initial fitting of several model forms, including neighbor-based models, tree-based models, support vector machines, and artificial neural networks. A neural network is selected as the most promising candidate and hyperparameter optimization using a genetic algorithm is performed. Finally, the resulting model, its potential applications, and areas for further research are discussed.

Preliminary Development and Verification of a Openfoam-Based Multi-Physics Coupling Safety Analysis Code for Lead-Based Reactor

Preliminary Development and Verification of a Openfoam-Based Multi-Physics Coupling Safety Analysis Code for Lead-Based Reactor

MELCOR integrated severe accident code application to safety assessment of high-temperature gas-cooled reactors

MELCOR is an integrated thermal hydraulics, accident progression and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (U.S. NRC) since the early 1980 s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application space to non-LWR reactor concepts. Current MELCOR development efforts have been focused on providing the U.S. NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC’s near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. In this paper, we describe the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. This model represents a TRi-structural ISOtropic particle fuel (TRISO) pebble bed HTGR with a primary system rejecting heat to a recuperative heat exchanger. Surrounding the reactor vessel is a reactor cavity contained within a confinement volume cooled by the Reactor Cavity Cooling System (RCCS). A range of demonstration calculations are performed to evaluate the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

Preliminary Development and Verification of Divertor Module for CFETR System Safety Analysis Code

China Fusion Engineering Test Reactor (CFETR), the next-generation fusion device of China, is designed to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and the commercial fusion power plant by solving the essential physical, engineering, and technical problems, including tritium self-sufficiency and high duty cycle of 0.3–0.5. Before the completion of the design and construction of CFETR, the thermal-hydraulic safety assessment is needed in order to ensure the safety of nuclear facilities and to get the national nuclear safety license. Up to now, there is no specialized code for the fusion reactor safety assessment. And the safety assessment work that had been done for the fusion device is completed by the system safety analysis code for pressurized water reactors, such as RELAP5, CHATHARE, and so on. For developing a simulation tool that can meet the safety assessment needs of the fusion reactor, some special models need to be established compared to the fission reactor. In this article, the development of a module for the divertor is presented. The thermal-hydraulic safety of the divertor with design basis accidents, and in-vessel loss of coolant (LOCA), is analyzed based on the developed code. Comparisons are performed to verify the reliability and feasibility of the developed code.

Scaled validation test for high prandtl number fluid mixed convection between parallel plates

The Kairos Power fluoride salt-cooled high temperature reactor (KP-FHR) passive decay heat removal system performance strongly depends on heat transfer characteristics in the reactor vessel’s down comer region. This region is an annular gap where the salt coolant flows downward between the core barrel and reactor vessel before entering the core region at the bottom of the vessel. Kairos Power’s principal safety analysis code, KP-SAM, extrapolates from existing heat transfer correlations for this geometry and applicable flow configurations. Main flow regimes of interest for safety analysis of the KP-FHR, when the system transitions from forced to natural circulation in the vessel upon loss of forced cooling, are the laminar and transition regimes. These regimes are not very well characterized in the literature, so in order to validate the inclusion of appropriate correlations in KP-SAM, scaled separate effects tests are conducted using a test facility that matches the Reynolds, Rayleigh and Prandtl numbers between the prototypical reactor coolant, Flibe, and a surrogate fluid, Therminol VP-1. The specific boundary conditions investigated are both symmetric heating and asymmetric heating. The data generated from this test serves to develop a validated heat transfer correlation for the laminar and transition regimes for implementation in KP-SAM, in support of safety analysis and licensing of the KP-FHR.