System Analysis Code Research Articles

Coupled irradiation-thermal-mechanical analysis of the solid-state core in a heat pipe cooled reactor

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027.3K for the fuel and 956.2K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

A Role of Interfacial Drag Force to Simulate Dam-Break Problem with Plant System Analysis Code, Trace

A Role of Interfacial Drag Force to Simulate Dam-Break Problem with Plant System Analysis Code, Trace

A Role of Interfacial Drag Force to Simulate Dam-Break Problem with Plant System Analysis Code, Trace

A Role of Interfacial Drag Force to Simulate Dam-Break Problem with Plant System Analysis Code, Trace

Application of different Krylov subspace methods in subcooled flow boiling simulation

In this work, a two-phase flow and heat transfer code was developed using the one-dimensional single-pressure six-equation two-phase flow model and the semi-implicit method. The developed code was validated with the Bartolomei’s subcooled flow boiling experiment and the numerical results indicated that the developed code has the capability for the simulation of subcooled flow boiling behavior. After that, the accuracy and efficiency performance of six Krylov subspace methods (CGS, BICG, BICGSTAB, GMRES, QMR and TFQMR) were evaluated. The results indicated that these six Krylov subspace methods have the capability for the accurate simulation of subcooled flow boiling behavior, and the TFQMR and CGS have better computational efficiency performance compared to other four Krylov subspace methods. This work could provide the benefit for the further development of nuclear reactor system analysis code.

Analysis of nodalization uncertainty for nuclear system analysis code with Lax-Wendroff numerical scheme

In modeling the complex systems such as a nuclear power plant (NPP), the nodalization becomes more important to satisfy Courant number limitation and reduce the numerical diffusion due to the 1st order numerical scheme. Since the higher-order numerical scheme can reduce the numerical diffusion problem, the accuracy can be improved and the nodalization uncertainty can be reduced by improving the current numerical scheme. However, another sources for the user effect uncertainty originates from a nodalization process and this can be affected by the selected numerical scheme. Sufficient sensitivity analyses should be performed on the nodalization to ensure that the calculated results are free from these uncertainties. Therefore, in this study, the nodalization uncertainties of 1st order upwind scheme and Lax-Wendroff scheme are evaluated with the revised MARS-KS code. The analysis of the nodalization uncertainty is performed for the separate effect tests (SETs) such as subcooled boiling and reflood experiments. The nodalization uncertainty is analyzed with the error between the experimental data and the calculation results of 1st order upwind scheme or Lax-Wendroff scheme. Thus, when using Lax-Wendroff scheme, the more accurate results with decreased nodalization uncertainty can be obtained. For the reflood calculations, accurate and consistent prediction of maximum peak cladding temperature (PCT) and accurate prediction of quenching time were observed with Lax-Wendroff scheme. However, the quenching time may not be consistent depending on the user. In terms of the computational efficiency, the central processing unit (CPU) time is generally increased by 40% with Lax-Wendroff scheme.

Development of a special thermal-hydraulic component model for the core makeup tank

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

Comparative study of constitutive relations implemented in RELAP5 and TRACE – Part II: Wall boiling heat transfer

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

Development of radiation characteristics analysis code system for geological disposal and application to vitrified waste disposal with various LWR burnup conditions

The radiation characteristics analysis code system for geological disposal is developed to study the radiation influence of high-level waste on an engineered barrier system for the various conditions from a reactor burnup through geological disposal. The verification of the code system is performed by the comparisons with the data in the previous study which analyzed the radioactivity, the radiation emission rate, the radiation energy spectrum, and the radiation flux for the vitrified waste fabricated from spent UO2 fuel with 45 GWd/t discharge burnup. The calculation of neutron emission rate and energy spectrum shows the improvement by the introduction of updated data, and the calculation results of the code system show the good agreement with the reported data, considering the calculation conditions. The effect of the burnup condition of spent nuclear fuel in a light water reactor (LWR) on the radiation characteristics of vitrified waste is surveyed by the application of the code system. The numerical analyses are performed on the basis of the burnup conditions of two types of fuels (UO2 and MOX) and two burnup levels of current standard (45 GWd/t) and high burnup (70 GWd/t). The ratio of high-level liquid waste in vitrified waste is 20 wt% for all cases. The analysis results indicate that the utilization of MOX fuel increases the MA radioactivity drastically, thereby increasing the neutron emission rate of waste and the neutron flux on the inner surface of the overpack by a factor of over nine in comparison to the UO2 fuel waste disposal, whereas the utilization of high burnup fuel increases the emission rate and flux of neutron and γ-ray by 20 – 170 %. The radiation fluence is sufficiently small even by the change of burnup condition compared to the lifetime fluence of reactor pressure vessel employed in LWR.

Assessing the effects of switching from multi-channel to three-dimensional nodalisations of the core in TRACE

Modern thermal-hydraulics (TH) systems analysis codes are increasingly being updated to include three-dimensional (3D) TH components towards capturing the complex flow phenomena in large components of light water reactors (LWRs). These are becoming increasingly popular for modelling reactor pressure vessels and cores since they potentially offer a more realistic representation of, for example, the effects of asymmetric loop flows and injection, which cannot be captured using traditional multi-channel models. Experience suggests, however, there may be unexpected side effects of upgrading from traditional multi-channel models. In this paper, we present comparative results demonstrating our experiences in switching to 3D components using the US-NRC system code TRACE. We present a series of studies in which 1D and 3D nodalisations are compared for representative PWR loss-of-coolant-accident (LOCA) scenarios. We see a clear trend for many of the LOCA scenarios considered. In particular, the peak cladding temperature (PCT) for 3D nodalisations is significantly lower than for traditional multi-channel models in many cases. Counter current flow limitation (CCFL) at the core upper tie plate appears to be one of the driving phenomena for many of the observed differences. This trend is, however, not universal; For a surge line break in the large scale test facility (LSTF), heterogeneities in the core upper tie plate lead to hot regions in the core, which are not captured in the multi-channel model. For large PWR cores, 3D nodalisations appear to represent the reality more closely. For smaller cross-sections, however, the validity of the TRACE’s CCFL prediction for 3D nodalisations is questionable. Specific guidelines are therefore needed to help TRACE users select the most appropriate 3D nodalisation for different geometries.

SCALE 6.2.4 Validation for Light Water Reactor Decay Heat Analysis

Energy release from the decay of radionuclides in nuclear fuel after its discharge from reactor is a critical parameter for design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. Well-validated computational tools and nuclear data are essential for decay heat prediction. This paper summarizes the validation of the SCALE nuclear analysis code system version 6.2.4, used with ENDF/B-VII.1 libraries, for decay heat analysis of light water reactor used fuel. The experimental data used for validation include full-assembly decay heat measurements that cover assembly burnups of 5 to 51 GWd/tonne U, cooling times after discharge in the 2- to 27-year range, and initial fuel enrichments up to 4 wt% 235U. The comparison between calculated (C) and experimental (E) decay heat showed very good agreement, with an average C/E over all considered measurements of 1.006 (σ = 0.016) for pressurized water reactor and 0.984 (σ = 0.077) for boiling water reactor assembly measurements. The effect of using assembly-average versus axially varying modeling data on the calculated decay heat, important to thermal analyses for used fuel transportation and storage systems, is discussed.

A modified system analysis code for thermo-hydraulic calculation of hydrogen in a nuclear thermal propulsion (NTP) system

Taking advantages of long lifetime and high power density, nuclear thermal propulsion (NTP) has become an advanced propulsion option in some special power fields. Due to the high heat release and dramatic density change, the flow instability and subsequent unbalance of flow and heat transfer would happen. The thermo-hydraulic calculation of working medium is an important part of the NTP system designing work, and the system analysis methods should be developed. In this study, the system analysis code RELAP5 was modified with the thermo-physical properties, flow and heat transfer correlations of the propellant hydrogen. The modified system code has been validated against the published experiment data. The calculation results show good agreement with the experimental results and the code ELM’s calculation results. The modified code was applied to thermo-hydraulic calculation of hydrogen in the analysis of a typical NTP system. The results indicate that the heat conduction between the inner and outer cores should be considered in the calculation of a two-pass NTP reactor core. This study has provided a reference for developing an NTP system analysis code.

Experimental research and model development on interfacial drag in rectangle channel bubbly and slug flow

As one of the most important constitutive models, interfacial drag determines the phase distribution directly and impose significant influences on pressure drop and heat transfer. Due to the limited study on interfacial drag for rectangle channel, experimental research and model development were performed in this paper. The air-water flow experiments in rectangular channels were carried out to obtain local parameters such as void fraction, interfacial area concentration and bubble chord length. These parameters were all characterized with core peak distribution except for the situations of void fraction at lower gas velocity and interfacial area concentration at higher gas velocity. The development of interfacial drag model in bubbly and slug flow was proposed based on the experimental data. The drag coefficient and interfacial area concentration in bubbly and slug flow were modified for distorted bubble, Taylor bubble and liquid slug respectively. Further validation was performed with system analysis codes. Better prediction of interfacial area concentration and void fraction, as well as numerical stability were obtained, proving the validity of interfacial drag model developed for rectangle channel bubbly and slug flow.

Study on Calculation Method of Courant Limit in Thermal Hydraulic System Analysis Code

In order to analyze the influence of different Courant limit calculation methods on the calculation speed and accuracy of reactor thermal hydraulic system analysis code, two kinds of Courant limit calculation methods were studied: synthesis method and grouping method. The calculation principles of the two methods were analyzed, and the two methods were adopted to calculate the steady-state condition and large break loss of coolant accident condition of a PWR respectively. The results show that there is no significant difference between the two methods in steady-state condition; in the case of large break loss of coolant accident with drastic change of velocity field, the synthesis method can obtain more accurate calculation results, but it takes more time, and the grouping method can obtain faster calculation speed, however its calculation accuracy is lower.

Performance analysis of PRHRS in primary and secondary circuit for offshore floating nuclear plant

In order to conduct performance analysis of passive residual heat removal system (PRHRS), a system analysis code applicable to offshore floating nuclear plant (OFNP) was developed based on existing code by adding motion condition model and the OFNP system with PRHRS in primary and secondary circuit was modeled in this study. The station blackout (SBO) accident was calculated under stationary and motion conditions with PRHRS in primary and secondary circuit operating respectively to compare the performance of the two PRHRS. The results show that the heat removal capability of PRHRS in secondary circuit is better than that in primary circuit with the same arrangement. The influence of heaving condition on the primary PRHRS is greater than the secondary PRHRS, while the secondary PRHRS is more susceptible to trimming and pitching conditions and heeling and rolling conditions have little influence on both of two PRHRS.

The development of ARSAC for modeling nuclear power plant system

A new safety analysis code for pressurized water reactors, ARSAC (Advanced Reactor System Analysis Code), is developed to satisfy the new demand for nuclear industry development in China. The code utilizes a hierarchical reactor system model to build the reactor model more conveniently, applies the general non-equilibrium non-homogeneous Euler-Euler six equation two-fluid models, and adopts a high-efficiency system solving algorithm. A semi-implicit scheme is applied to the conservation equation and a NRLU algorithm is implemented. The code structure of the ARSAC is highly modular with C++ for further improvement. The preliminary code assessment on conceptual, separate effect, and integral effect experiments showed that the prediction is reasonable. The ARSAC code will be used for safety analysis of operating PWRs and the design of innovative reactors including small modular reactors.

A three-dimensional PWR LBLOCA simulation using the CUPID code

As the need for simulations of multi-dimensional behavior in the safety analysis is increasing, efforts have been devoted to developing a three-dimensional module for the conventional system analysis codes. After having developed a one-dimensional system analysis code, MARS, the Korea Atomic Energy Research Institute (KAERI) began to develop the CUPID code to address the need for a multi-dimensional analysis. Recently, the reactor vessel module, such as a package of interfacial heat and momentum transfer model, wall friction and wall-to-fluid heat transfer model, were implemented in the CUPID and named CUPID-RV module. The final goal of this study was an analysis of the Large Break Loss of Coolant Accident (LBLOCA) at a rod-by-rod level, and the study focused on the validation of LBLOCA thermal-hydraulics phenomena and the application to a three-dimensional LBLOCA simulation. In this paper, we introduce the validations using Edward pipe, UPTF, and FLECHT-SEASET experiments related to the LBLOCA blowdown, refill, and reflood phases and a three-dimensional LBLOCA simulation for the APR1400 PWR reactor pressure vessel using simple boundary conditions at an assembly-by-assembly level as an application.

Development and verification of fast reactor burnup calculation module FRBurner in code system CBZ

ABSTRACT CBZ is a general-purpose reactor physics analysis code system, and FRBurner, which focuses on fast reactor burnup calculations, was developed recently with diverse combinations of available methodologies. Verification of this module is conducted with the OECD/NEA fast rector benchmark since this benchmark provides various types of fast reactors. Four key reactor physics parameters, effective neutron multiplication factor k eff, effective delayed neutron fraction β eff, sodium void reactivity ∆ρ void, and Doppler reactivity ∆ρ Doppler are the focus and compared to two references provided by JAEA and CEA, respectively. The biases between the results from FRBurner and the JAEA and CEA references on each of the above key parameters are less than 0.5%, 1%, 3% and 7%, and less than 1.0%, 4%, 12%, and 12%, respectively. The comparison indicates that the FRBurner module would provide acceptable results for general-type fast reactor physics analysis in research. As one innovation, the detailed burnup chain model, which is significantly different from a generally used pseudo fission product model in fast reactor neutronic analysis, is applied in FRBurner. The detailed burnup chain model helps FRBurner explicitly provide information about the inventory of fission products for nuclear waste management and spent fuel reprocessing.

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.

Comparative study with two reduced test facilities for air-cooled RCCS scaling law

Korea Atomic Energy Research Institute (KAERI) and Argonne National Laboratory (ANL) conducted steady-state tests at constant heat flux conditions to investigate the scale effect on heat removal behavior of the air-cooled Reactor Cavity Cooling System (RCCS). Two differently scaled-down test facilities were used in this study, KAERI’s ¼ scale with 4.0-m cavity height, and ANL’s ½ scale with 6.8-m cavity height. Two scaling laws were proposed to simulate the radiation across the cavity and the buoyancy-driven duct flow in the riser, respectively. The test matrix for KAERI and ANL test facilities focused on the scaling effect of the heated riser length with constant heat flux at the PMR200 RCCS design. The test results showed that mixed convection in the riser duct is an important factor for accurately extrapolating the thermo-fluid behavior in the prototype from the test results in the scale-down facilities. The system analysis code, GAMMA+, with improved heat transfer models predicted fairly well the air-cooled RCCS test data from two facilities. GAMMA+ analysis results showed that the predicted radiation fractions on the heated plate in the scale-down test conditions were larger than those on the reactor vessel in the prototype. The scaling law for air-cooled RCCS was improved by considering the mixed convection in the riser duct and the same radiation fraction on the heated plate. The mixed convection effect was calculated by the height ratio. The same radiation fraction provided a conservative extrapolation from the scale-down test results.

Dynamic simulation of the gas-cooled space nuclear reactor system using SIMCODE

One dimensional space nuclear reactor system analysis code—SIMCODE is developed and five typical transient operation conditions are calculated, including the system startup, the loss of flow accident, the loss of heat sink accident, the reactivity insertion accident, and the small loss of coolant accident. These conditions are part of the key performance of the space nuclear reactor system and are related with both the individual characteristics of each component and the coupled characteristics between all the components. By analyzing the simulation results, system transient performance and coupled system characteristics are acquired. The information can provide helpful information for the system evaluation, safety system design, and emergency strategy setup. Based on the simulation, system control suggestions are also proposed and will be beneficial for further researches.