System Analysis Code Research Articles

Simulation investigation of ocean motion on small reactor based on modified system code

Offshore nuclear small reactor is a promising research direction during recent years. In order to carry out system simulation and safety analysis under accident conditions for marine reactors, the classic thermal-hydraulic system analysis code RELAP5 is modified by taking heaving, inclining and rolling motion into consideration in this paper. Additional forces are added to the momentum equation for heaving and rolling, and control volume coordination are transformed for inclining and rolling. Then, a simplified small reactor with two loops is simulated by the modified code and the effect of ocean motion is investigated. Generally, heaving leads to the periodic fluctuation of mass flow rate, inclining would cause the decrease of total mass flow through the core, and the effect of rolling is more complex. Simulation analysis is also performed on two design basis accident LOCA and LOFA in this study. Safety related conclusions are obtained about the impacts of various ocean motion.

Best estimate plus uncertainty analysis of the China advanced large-scale PWR during LBLOCA scenarios

China advanced large-scale Pressurized Water Reactor (PWR) implements series of passive safety systems to provide core cooling and decay heat removal in case of some unexpected accidents. The development of reactor best estimate system analysis codes has realized the transition from conservative safety analysis to Best Estimate plus Uncertainty (BEPU) analysis for the purpose of achieving more realistic results and digging larger safety margins. The BEPU analysis during Large Break Loss of Coolant Accident (LBLOCA) has been widely used for the licensing application of Nuclear Power Plants (NPPs). In this paper, the BEPU analysis of China advanced large-scale PWR was performed under the conditions of LBLOCA scenarios employing RELAP5 code. The models of reactor coolant system with all the passive safety systems were established and the influence of uncertainties of important parameters on the Peak Cladding Temperature (PCT) was evaluated. The results show the obtained PTC (1314 K) through the BEPU analysis is far below the 10CFR50.46 PCT acceptance criteria (1477 K) in case of LBLOCA and is also lower than the conservative calculation value (1348 K), indicating a larger safety margin. This work provides a meaningful evaluation method for China advanced large-scale PWR safety assessment.

Research on code coupling method in severe accident simulator development

After Fukushima nuclear accident, the newly published nuclear regulation featured requests for the development of severe accident (SA) simulators in operator training. In response to such requests, this research deals with a code coupling method, the coupling of thermal-hydraulic system analysis code with integral SA code. Based on the existing full-scope simulator (FSS), the architecture of SA simulator was designed through the method. The crucial technical route is that the RELAP5/MOD3.2 code is utilized for early response, when the coupling condition is met, it switches to MAAP4.0.3 code to calculate middle-late accident. To ensure smooth transition, their major differences were analyzed and some coupling parameters were selected. A test case of SBLOCA was performed via RELAP5-couple-MAAP4, and some specific quantitative assessments were defined to justify the method. The results show a smooth transition between RELAP5 and MAAP4 is achieved and is expected to extend FSS function to SA field.

Derivation of bootstrap current fraction scaling formula for 0-D system code analysis

We constructed new scaling formulas for the bootstrap current faction fBS for tokamaks with full non-inductive operation with positive shear safety factor q profiles. A fitting database was built by the ACCOME code, which contains 12,150 cases of self-consistent equilibria with the q profiles satisfying qρ≥1.0 and q95≥3.0 in 0≤ρ<1, where ρ is the normalized minor radius and q95 is q at the 95% poloidal flux surface. From the fitting database, we derived the formulas by the multiple regression analysis. The best formula contained the inverse aspect ratio ε, the poloidal beta, the internal inductance, q95/q0, and the profile shape indices for density and temperature as input. The Root Mean Squared Error for the fitting database of this formula was 0.025. We tested the derived formulas to three kinds of datasets and compared with the K. Gi models and the ITER Physics Design Guidelines model. The first test is for the compact neutron source with the same plasma size with the fitting database, the second (including reversed shear) and the third ones (positive shear only) are for the conventional reactor size. The best formula seems to be applicable to tokamak plasmas with a wide range of parameters.

Application of adjoint based node optimization method to nuclear thermal-hydraulic system analysis code

Nuclear thermal-hydraulic system analysis codes with the Best Estimate plus Uncertainty approach are unavoidably uncertain and naturally, an uncertainty analysis has been an important issue. In general, the system is discretized depending on the user while satisfying some criteria. Previously, uncertainties related to spatial discretization in nodalization have been less considered due to their difficulty in quantification. However, there are some results that they are comparable to other uncertainties. Previous studies have been conducted by evaluating and optimizing the uncertainty related to user effect by performing iterative calculations. Since these methods consume a large amount of resources for computation, this study suggests a method of consuming less resources. For this purpose, the adjoint method is applied to the code for node optimization in this study. The proposed methodology is applied to two cases and confirmed its applicability for the node optimization and its utility in the node uncertainty quantification.

Node configuration uncertainty in nuclear safety analyses

In nuclear safety analyses, a 1-D thermal hydraulic system analysis code with the best estimate plus uncertainty (BEPU) approach is generally used. There are sources of uncertainty such as nodalization, physical models, and numerical scheme. Among them, the authors specifically focused on the node configuration uncertainty related to spatial discretization that is regarded as less dominant sources of uncertainty in the code previously. Furthermore, it was not easy to quantify the uncertainty from the node configuration. The authors quantified and analyzed the node configuration uncertainty to understand how much it affects the solution. In this study, the effect of spatial discretization are studied by discussing simulation results for various spatially discretized node systems. The discussion on the node configuration uncertainty will be presented by examples of SUbcooled BOiling (SUBO) test simulation and Loss Of Fluid Test (LOFT) simulation using Multi-dimensional Analysis of Reactor Safety (MARS) code developed in South Korea. From SUBO test simulation, it was found that when the number of nodes is small than the node configuration uncertainty can overwhelm the uncertainty from the physical model. From LOFT simulation, it was revealed that the node configuration can cause bifurcation in the complex nuclear system analyses. Thus, it is concluded that the node configuration uncertainty can be quite significant and has to be considered in the nuclear system analyses.

System code analysis of accumulator nitrogen discharge during LOCA experiment at PWR PACTEL test facility

The hydroaccumulators in pressurized water reactors can inject nitrogen into the reactor system. In the primary system, nitrogen affects core cooling and accident management, both adversely and beneficially. The PWR PACTEL experiment NCG-13 have shown that during a hot leg SB LOCA, nitrogen in the primary side can block the primary to secondary heat transfer and thereby prevent the primary depressurization to the point needed for the long-term accident management. This paper presents the APROS and TRACE calculations of the PWR PACTEL NCG-13 experiment. Both codes calculate the transient progression and the timing of the main events satisfactorily, once the suitable options are selected and adjustments made. There is, however, one big discrepancy between the code simulations and the experiment: in the simulations, much more nitrogen is needed to get qualitatively the same behaviour. The difference is a factor of 2.5 by mass for stopping the depressurization and 4–6 to cause a core heat-up. This is of concern regarding the confidence in the codes, as the simulations underestimate the adverse effect of nitrogen on the core coolability.

A multi-scale and multi-physics approach to main steam line break accidents using coupled MASTER/CUPID/MARS code

The focus of this study is on a multi-scale and multi-physics nuclear safety analysis method for main steam line break (MSLB) accidents. The indirect coupled code MASTER/CUPID using MASTER DLL was combined directly with MARS to form MASTER/CUPID/MARS. Whereas MASTER is a three-dimensional neutronics code, CUPID is a three-dimensional thermal-hydraulics code and MARS is a one-dimensional nuclear system analysis code. The coupled MASTER/CUPID/MARS code is applied to the analysis of a Pressurized Water Reactor (PWR) MSLB accident, and the calculation indicates that the proposed four-step method for the CUPID/MARS multi-scale analysis is also valid for the multi-scale and multi-physics safety analysis using the MASTER/CUPID/MARS code. It can also provide enough information to analyze the nuclear safety of such conventional time-dependent variables for 1-D reactor components and the three-dimensional temperature distribution for the 3-D reactor pressure vessel.

Control and thermal analysis for SCWR startup

Startup system, the design of startup sequences analysis is an important part of SCWR design. A thermal hydraulic system analysis code for supercritical water reactor named SCTRAN is used to model the entire startup system based on the circulation loop for startup and once-through direct cycle. The problem of the heat transfer coefficient (HTC) does not accurately capture deterioration phenomenon, the HTC is calculated as a discontinuity in the mode transfer region, its low prediction accuracy above the quasi-critical region have been solved by the new wall heat transfer model. Especially, the look-up table would not be used to obtain the HTC and achieves high prediction accuracy across the critical region, unless the pressure is higher than 19 MPa. After that, to get a smooth recirculation variable pressure startup process, the system model integrates the control system which can controls the temperature, the steam drum water level, the thermal power, and the coolant flow rate. Based on the CSR1000 core and entire once-through direct cycle and circulation loop for startup, four stages under control systems, from low pressure to full power condition in recirculation startup process, were analyzed with code SCTRAN and wall heat transfer model was modified. The calculation results show that the recirculation system can startup from subcritical state to full power state without issue of CHF. The control system can control the parameters quite well and maximum cladding temperature (MCST) can be limited under 650 °C in the startup process. The modified SCTRAN code in this paper can further expand the computational range and computational accuracy. The full-scale control system can meet the needs of parameters expected response.

Steady states and LOFA analyses of the updated WCCB blanket for multiple fusion power modes of CFETR

The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates of Chinese Fusion Engineering Test Reactor (CFETR). The core parameters of CFETR have been updated to major radius R = 7.2 m, minor radius a = 2.2 in 2018. Accordingly, the WCCB blanket design was also updated to satisfy multiple operation modes of CFETR with the fusion power of 200 MW, 500 MW, 1.0 GW and 1.5 GW. The WCCB blanket is characterized by 3 independent cooling systems (CSs), namely CS1 for the cooling of the First Wall (FW), CS2&3 for Breeding Zones (BZs). At 200 MW of low fusion power, CS3 is idle. At 500 MW to 1.5 GW of high fusion power, all 3 CSs are put into use to ensure adequate cooling. So it is indispensable to evaluate the feasibility of the updated WCCB blanket from the thermal hydraulic and safety point of view. A system analysis code RELAP5, was employed to model the blanket. The steady states were analyzed under 200 MW, 500 MW, 1.0 GW and 1.5 GW of fusion power. Then, the Loss of Flow Accident (LOFA) was simulated with different assumptions to verify the robustness of 3 cooling systems design under different initial postulations. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature, can basically meet the design criterion, which preliminarily verifies the feasibility of the WCCB blanket from the thermal hydraulic safety point of view. Comprehensive system design and analysis is under way.

Analysis of the SPERT-III E-Core experiment using CASMO5/TRACE/PARCS based on JENDL-4.0 and ENDF/B-VII.1

ABSTRACT An analysis of the Special Power Excursion Test III E-Core experiment was performed in order to confirm the calculation accuracy of the light-water-reactor core analysis code system, constructed by the authors, of the CASMO5 and the TRACE/PARCS for the reactivity-initiated accident (RIA) analysis. The influence of the resonance up-scatter model (RUM) and the effective Doppler temperature model (EDTM) in the CASMO5 on the Doppler reactivity feedback effect was also discussed through the comparison with the conventional calculation without those models and the perturbation calculation of the Doppler reactivity coefficient. The calculation results by the CASMO5/TRACE/PARCS mostly showed good agreement with the experimental data within the range of experimental uncertainty, which confirmed the calculation accuracy of the analysis code system for the RIA analysis. In the calculation, the JENDL-4.0 and the ENDF/B-VII.1 were used, however, obvious difference was not seen in the calculation results between the two nuclear data libraries. The influence of the RUM on the core parameters was less than that of the 10% increment of Doppler reactivity coefficient on the conventional calculation. The influence of the EDTM became large on the cold condition, which was the same tendency on the Doppler reactivity coefficient discussed in the previous studies.

Verification and improvement of dynamic motion model in MARS for marine reactor thermal-hydraulic analysis under ocean condition

Abstract Unlike land-based nuclear power plants, a marine or floating reactor is affected by external forces due to ocean conditions. These external forces can cause additional accelerations and affect each system and equipment of the marine reactor. Therefore, in designing a marine reactor and evaluating its performance and stability, a thermal hydraulic safety analysis code is necessary to consider the thermal hydrodynamic effects of ship motion. MARS, which is a reactor system analysis code, includes a dynamic motion model that can simulate the thermal-hydraulic phenomena under three-dimensional motion by calculating the body force term included in the momentum equation. In this study, it was verified that the dynamic motion model can simulate fluid motion with reasonable accuracy using conceptual problems. In addition, two modifications were made to the dynamic motion model; first, a user-supplied table to simulate a realistic ship motion was implemented, and second, the flow regime map determination algorithm was improved by calculating the volume inclination information at every time step if the dynamic motion model was activated. With these modifications, MARS could simulate the thermal-hydraulic phenomena under ocean motion more realistically.

Impact of system parameters on quenching heat transfer in the candidate materials for accident tolerant fuel-cladding in LWRs

We investigated a potential impact of the candidate materials (FeCrAl, SiC-CVD, and Zry-4) for the Accident Tolerant Fuel-cladding (ATF) in the quenching heat transfer according to the system parameters such as liquid subcooling, pressure, and flooding rates. Results obtained by water quenching experiment and system code analysis using MARS 3D-VESSEL are presented that the quenching heat transfer including the post-CHF parameters (i.e., minimum film-boiling temperature) is more influenced by the system parameters compared to change of the cladding materials. Current models of minimum film-boiling temperature in system code are also discussed for the application of ATF and its oxidation. Select option in the model library is proposed to consider the surface parameters using weighting term in thermal-hydraulic system code.

SPARK-LS: A novel design of a compact liquid chloride salt cooled fast reactor based on tube-in-duct fuel concept

A novel design of a compact liquid chloride salt cooled fast reactor (SPARK-LS) was proposed in the present study as a distributed energy resource for easy transportation and convenient deployment in diverse applications, such as electricity generation and district heating. The SPARK-LS reactor with rated power of 40 MWth can operate for 10 effective full power years without refueling. An innovative tube-in-duct fuel concept with coolant inside the tubes and fuel outside was adopted to achieve higher fuel volume fraction and lower pressure drop when compared with traditional pin-type fuel. Neutronics and thermal-hydraulics design and preliminary safety analysis were accomplished to obtain an optimized core design by using the advanced reactor design and analysis code system SARAX. Titanium was adopted as the reflector material due to its low density, good corrosion resistance, large strength-to-density ratio and decreased albedo with the neutron spectrum hardening. The optimized core design featured with compact volume and light weight was proved to satisfy the design constraints of peak coolant temperature, peak fuel centerline temperature and coolant pressure drop. The negative coolant density reactivity coefficient in the full core was ensured during the whole core lifetime thanks to the compact core design and the adoption of the titanium reflector. Based on the quasi-static reactivity balance method, five anticipated transients without scram were analyzed to confirm the inherent safety feature of SPARK-LS. It was shown that passive shutdown could be achieved safely for all the transients and the residual decay heat after the shutdown could be removed automatically by the natural circulation of coolant.

Thermal hydraulic model of the molten salt reactor experiment with the NEAMS system analysis module

System analysis codes have a long history of providing best-estimate and conservative safety analysis for both light water and advanced reactor technologies, including molten salt reactors. As interest continues to expand with advanced reactor concepts, system analysis codes will need revisions to accommodate the behavior of these technologies. Legacy system analysis codes will need to be updated to the latest numerical techniques to shorten execution time and increase the accuracy of results. One example of a modern system analysis code that already encompasses these characteristics is the System Analysis Module (SAM).One key objective of this paper was to review available information for system code modeling of the Molten Salt Reactor Experiment (MSRE) from sources in the open literature and collect the information from these open sources in one place for the first time. This supports the potential objective of developing an open specification for system code analysis for MSRE steady state and transients with and without reactor kinetics. Data from actual MSRE tests will serve as the basis for code-to-code comparison exercises, including the MSRE zero power physics tests, the fuel pump start-up and coast down tests, and the natural circulation transient. The objective is to produce a code-to-code benchmark with a standardized set of comparison problems, recognizing the limitations of the original data.To demonstrate an initial application of this objective and the usefulness of compiling this open data, two Molten Salt Reactor Experiment (MSRE)-related models were developed to evaluate SAM for liquid fueled molten salt reactors. One model was the SAM MSRE hydraulic mockup, which provided experimental data for pressure drop measurements. The second model was the complete MSRE primary loop. The MSRE primary loop model incorporated a fluoride salt fuel/coolant with heat transfer in both the core and heat exchanger. For both the hydraulic mockup and MSRE primary loop models, a holistic 1-D system description was built using open documentation, an open description that can be readily modified and applied for any system analysis code.SAM results for the pressure drop of the hydraulic mockup model were within 6% with measurements. Coolant temperatures for the primary loop model matched the expected axial change in temperature from historical calculations. Using alternative coolant properties obtained from the literature, corresponding to salts with different actinide contents, returned similar trends in core temperature profiles. A thermal hydraulic demonstration of a loss-of-flow transient showed the importance of coupling SAM thermal hydraulic analysis to neutronics. This coupling is essential for simulating MSR transients with system analysis codes.

Numerical approach to study the thermal-hydraulic characteristics of Reactor Vessel Cooling system in sodium-cooled fast reactors

The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

Improvement of RELAP5 code for density wave instability analysis in parallel narrow rectangular channels

As one of the most common types of oscillatory instability, density wave oscillation (DWO) is of practical importance in the general analysis of two-phase boiling systems. Due to the confinement effect of narrow space, the two-phase phenomena in thin rectangular channels are probably different from those in conventional pipes with large diameters. As the widely accepted system analysis code, RELAP5 has the capability to simulate the DWO phenomena in narrow channels, but the accuracy is limited under the condition of different pressures, flow rates, or subcoolings. In this paper, modifications for related constitutive models are proposed based on single-variable sensitivity analysis. Two-phase friction model and nucleate boiling heat transfer model are selected as the most effective models to improve the predictability of RELAP5. The influence of both models mentioned above is analyzed and it reveals that they are related to two-phase pressure drop and the vapor void fraction distribution along the channel, respectively. The modified calculation is assessed against the experimental results under the wide range of pressure 1–10 MPa, and these comparisons are performed in the map of stability boundary with two dimensionless parameters, phase change number Npch and subcooling number Nsub. Then the heating power and the period of flow oscillation under threshold condition are assessed in detail. The results show that a better agreement can be obtained between the modified calculations and the measured experimental data.

Accident safety analysis of flow blockage in an assembly in the JRR-3M research reactor using system code RELAP5 and CFD code FLUENT

The main purpose of the paper is to execute the thermal-hydraulic analysis of flow blockage in a single fuel assembly in the JRR-3M research reactor (Upgraded Japan Research Reactor No. 3). Using the one-dimensional system analysis code RELAP5/MOD3.4, flow region in a standard fuel assembly (SFA) is lumped into a coolant channel. 5 similar coolant channels representing 5 independent assemblies are modeled to consider the interaction between blocked channel and adjacent channels. The rest of the core is lumped into 1 average channel and 1 bypass channel. The coolant system is also modeled in detail. Meanwhile, a three-dimensional model of heated part in the assembly is built using the computational fluid dynamics (CFD) code FLUENT. Results calculated by RELAP5/MOD3.4 are used as the boundary conditions of the 3D model. The user-defined function (UDF) is adopted to describe phase change and the power distribution in axial and radial direction in the assembly. Results indicate that it is necessary for assembly blockage to consider the influence of power distribution on accident consequence. When the blockage ratio is 64%, coolant in the hottest subchannel is still supercooled while coolant temperature at outlet is close to saturation temperature. It is the critical blockage ratio to ensure the reactor safety. When the blockage ratio is 70%, departure from nucleate boiling (DNB) will occur. According to the FLUENT code, the process of bubble generation and growth is discussed. It can be found that if the bubbles largely generate, there will be obvious impacts on heat transfer of fuel plate and coolant flow. Fuel plates may damage locally.

Reflood experiments at elevated pressures using intact and deformed rod bundles to simulate small and medium break loss-of-coolant accidents

Abstract Reflood experiments at elevated pressures for small and medium break loss-of-coolant accidents were performed to investigate the influence of fuel deformation on the peak cladding temperature and local quenching behaviors. The typical experimental parameters such as initial water level, power, reflood rate, system pressure, initial wall temperature, and fluid temperature were varied to cover a wide range of flow conditions under a medium and high pressure. The experimental results showed remarkable differences in the peak cladding temperature and local quenching behaviors between the intact and deformed rod bundles due to the flow blockage effect. The present results implied that the fuel rod deformation can enhance the coolability of fuel rods by the flow blockage in the present fully blocked channel. Any bypass in a partially blocked channel may change those results due to a crossflow between intact and deformed rods. The experimental results were then used to validate the system analysis code, Multi-dimensional Analysis of Reactor Safety-KINS Standard (MARS-KS). The MARS-KS code showed good prediction of the present experimental data for most of the flow conditions.

Demonstration of LOTUS multiphysics BEPU analysis framework for LB-LOCA simulations

The LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Labs (INL) is a plug and play, multiphysics environment to be used in support of system level plant transient analysis. New proposed rule changes in LOCA safety regulations (10 CFR 50.46c) require the inclusion of cladding hydrogen content in the evaluation of equivalent cladding reacted (ECR) and peak cladding temperature (PCT). The potential for a future implementation of said rule change motivates the current development of a unique union of core design, system analysis, and fuel performance codes within a framework allowing for uncertainty quantification (UQ). A demonstration of LOTUS capabilities to address this potential need is performed with an integration of the core design code PHISICS, the fuel performance code FRAPCON, and the system analysis code RELAP5-3D. UQ is performed via Wilks’ method to compute the one sided 95/95 confidence values of the aforementioned safety metrics. Results demonstrate the benefits of the Multiphysics Best Estimate Plus Uncertainty (MP-BEPU) methodology and provide useful visualization of the limiting cases.