Maximum Fuel Centerline Temperature Research Articles

CTF-PARCS Core Multi-Physics Computational Framework for Efficient LWR Steady-State, Depletion and Transient Uncertainty Quantification

Best Estimate Plus Uncertainty (BEPU) approaches for nuclear reactor applications have been extensively developed in recent years. The challenge for BEPU approaches is to achieve multi-physics modeling with an acceptable computational cost while preserving a reasonable fidelity of the physics modeled. In this work, we present the core multi-physics computational framework developed for the efficient computation of uncertainties in Light Water Reactor (LWR) simulations. The subchannel thermal-hydraulic code CTF and the nodal expansion neutronic code PARCS are coupled for the multi-physics modeling (CTF-PARCS). The computational framework is discussed in detail from the Polaris lattice calculations up to the CTF-PARCS coupling approaches. Sampler is used to perturb the multi-group microscopic cross-sections, fission yields and manufacturing parameters, while Dakota is used to sample the CTF input parameters and the boundary conditions. Python scripts were developed to automatize and modularize both pre- and post-processing. The current state of the framework allows the consistent perturbation of inputs across neutronics and thermal-hydraulics modeling. Improvements to the standard thermal-hydraulics modeling for such coupling approaches have been implemented in CTF to allow the usage of 3D burnup distribution, calculation of the radial power and the burnup profile, and the usage of Santamarina effective Doppler temperature. The uncertainty quantification approach allows the treatment of both scalar and functional quantities and can estimate correlation between the multi-physics outputs of interest and up to the originally perturbed microscopic cross-sections and yields. The computational framework is applied to three exercises of the LWR Uncertainty Analysis in Modeling Phase III benchmark. The exercises cover steady-state, depletion and transient calculations. The results show that the maximum fuel centerline temperature across all exercises is 2474K with 1.7% uncertainty and that the most correlated inputs are the 238U inelastic and elastic cross-sections above 1 MeV.

Performance analysis of nuclear reactor core loaded with Accident-Tolerant Fuel: Mo/Cr metallic microcell UO2 pellets and CrAl coating

This paper presents a performance analysis of a commercial pressurized water reactor (PWR) core loaded with accident-tolerant fuel (ATF) containing Mo/Cr metallic microcell UO2 pellets and a CrAl coating on the cladding. Metallic microcell UO2 pellets using additive materials, such as Mo and Cr, with coated cladding have been developed by the Korea Atomic Energy Research Institute (KAERI) for an OPR-1000 reactor. As demonstrated by the fuel pellet developed at KAERI, a metallic microcell UO2 pellet can increase the thermal conductivity of fuel pellets by forming a metallic film that encapsulates the UO2 to connect the metallic microcell and UO2 pellets; however, there has been no systematic core performance and safety analysis of the core with ATFs. In this study, core analysis with ATFs was conducted from a neutronic perspective. 1) The fuel assembly (FA) containing Mo/Cr metallic microcell UO2 pellets was analyzed using the lattice code STREAM developed by Ulsan National Institute of Science and Technology in terms of the Mo depletion chain and the impact of resonance treatment. 2) The initial and equilibrium cores were analyzed using STREAM/RAST-K 2.0 with respect to the impact of the additives and CrAl coating on the cycle length, power peaking, and fuel temperatures, etc. The reductions in the equilibrium core cycle length were 61, 44, and 65 effective full power days for each material type. 3) Finally, a rod ejection accident (REA) analysis was performed on the core using a Mo metallic microcell UO2 pellet and CrAl-coated ATF. Moreover, the maximum fuel centerline temperature of the ATF core was determined to be approximately 200 ℃ lower than that of the reference UO2 core. Furthermore, the presence of additional thermal margins in ATF cores for steady-state and accident transient scenarios was verified through a systematic neutronic analysis, and future studies will focus on the optimization of the ATF-loaded core by exploiting the extra margins achieved by ATFs and increasing the cycle length.

Sensitivity analysis of in-pile CHF experiments in the TREAT facility: Characterization of impacts of fuel system thermal properties

Understanding fuel-to-coolant heat transfer mechanisms under transient heating irradiation conditions is an important aspect of pressurized-water reactor (PWR) safety. The work here aims to elucidate the potential impacts of critical heat flux (CHF) events on nuclear fuel systems, utilizing the heat transfer time constant (HTTC) as the fundamental basis. A set of thermophysical material and heat transfer coolant properties was established to represent the HTTCs of two Transient Reactor Test (TREAT) facility experimental fuel/cladding designs. Pool and flow boiling conditions were used to model these fuel systems within TREAT experimental capabilities, and a power pulse was used to represent a transient insertion of energy. Using Sobol sensitivity analysis and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code, an investigation was conducted to identify the most important HTTC parameters pertaining to the output of the peak outer cladding temperature (POCT), maximum fuel centerline temperature, and the time of CHF.The maximum fuel centerline temperature was shown to be highly dependent on the UO2 volumetric heat capacity. Regarding the POCTs, DNB was determined to have a dominant effect on the heat transfer mechanisms of these fuel systems. Further, the HTTC significantly impacted POCT predictions with the UO2 volumetric heat capacity, the film boiling heat transfer coefficient, and the CHF multiplier being most influential regardless of coolant flow conditions. Gap variations were shown to influence the POCT in cases that experienced the DNB event. In non-DNB cases, the HTTC had no significant impact on POCT. For the time of CHF, variations in gap thickness, and the CHF multiplier were overwhelmingly dominant.

Analysis of neutron physics and thermal hydraulics for fuel assembly of small modular reactor loaded with ATFs

The new small modular reactor (SMR) NuScale has been in the approval process for Nuclear Regulatory Commission by its off-site fabrication and assembly and good safety performance. To ensure the safety of the fuel rod the rod ejection accident at the beginning of cycle (BOC) and the end of cycle (EOC), the traditional fuel rod is replaced by accident tolerant fuels (ATFs). Neutron physics and thermal hydraulic analysis need to be performed before the actual application to ensure feasibility. In this paper, neutron physics and thermal hydraulics for SMR loaded with ATFs are analyzed. The results show that the 4% uranium enriched ATF pellets can meet the lifetime requirements, and the application of silicon uranium fuel pellets can significantly reduce maximum fuel centerline temperature (MFCT). Safety margin analysis was also performed for different ATF material failure criteria. The results show that ATF materials can enhance the safety of NuScale reactors.

Research on thermal behaviors of several accident tolerant fuels based on 5 × 5 bundle subchannel model

Accident tolerant fuel (ATF) materials need to be studied by subchannel analysis under a variety of transient conditions before being applied to Light-Water-Reactors (LWRs). The present study assesses the thermal behaviors of various ATF claddings and fuels under four different transient conditions which will give rise to the occurrence of Departure from Nucleate Boiling (DNB) phenomenon. A 5 × 5 rod bundle model is built to conduct the subchannel analysis, and maximum cladding temperature (MCT) and maximum fuel centerline temperature (MFCT) of different ATF cladding and fuel groups are obtained. The simulation results show that both MCT and MFCT are reduced with the utilization of FeCrAl cladding, superior to HNLS/ML-A and SA3/PyC150-A claddings that only reduce MCT but increase MFCT. Significant MFCT reduction but slight MCT increase are achieved in most transient cases with the application of UO2 + BeO/SiC fuel pellets. Due to the different safety limit of various ATF materials, safety margin analysis is performed as well. The results show FeCrAl possibly is the best alternative cladding material among claddings studied, because both cladding safety margin (CSM) and fuel pellet safety margin (FPSM) are improved. Compared to single UO2 fuel, a larger fuel pellet safety margin is still provided by UO2 + BeO/SiC fuel pellets.

Application of regression, variance, and density based global sensitivity methods to integrated VERA-CS and BISON simulations

Uncertainty quantification (UQ) and sensitivity analyses (SA) are performed for coupled simulations between VERA-CS, a coupled pin resolved neutron transport and subchannel thermal hydraulics code, and the fuel performance code BISON. The interface between VERA-CS and BISON is performed in a multiphysics environment known as the LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Laboratory (INL). A focus is placed on using a variety of SA measures, including two regression based (Pearson and Spearman), one variance based (Sobol indices), and three moment independent measures (Delta moment independent measures with L1, L2, and L∞ norms). The problem under inspection is a single assembly depletion case for three fuel cycles. The figures of merit are the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). SA results show MDNBR to be linear with consistent rankings throughout the fuel cycles. MFCT is linear, but with a change in rankings at the switch from open gap to closed gap models. GCPP is nonlinear at intermediate states that coincide with the onset of contact between fuel and cladding. These nonlinear states allow for the showcasing of higher order SA measures over first order methods.

Sensitivity analysis of VERA-CS and FRAPCON coupling in a multiphysics environment

A demonstration and description of the LOCA Toolkit for US light water reactors (LOTUS) is presented. Through LOTUS, the core simulator VERA-CS developed by CASL is coupled with the fuel performance code FRAPCON. The coupling is performed with consistent uncertainty propagation with all model inconsistencies being well-documented. Monte Carlo sampling is performed on a single 17×17 fuel assembly with a three cycle depletion case. Both uncertainty quantification (UQ) and sensitivity analysis (SA) are used at multiple states within the simulation to elucidate the behavior of minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). The SA metrics used are the Pearson correlation coefficient, Sobol sensitivity indices, and the density-based, delta moment independent measures. Results for MDNBR show consistency among all SA measures, as well for all states throughout the fuel lifecycle. MFCT results contain consistent rankings between SA measures, but show differences throughout the lifecycle. GCPP exhibits predominantly linear relations at low and high burnup, but highly nonlinear relations at intermediate burnup due to abrupt shifts between models. Such behavior is largely undetectable to traditional regression or variance-based methods and demonstrates the utility of density-based methods.

A COMPARISON IN THERMAL-HYDRAULICS ANALYSIS OF PWR1000 USING FIXED AND TEMPERATURE FUNCTION OF THERMAL CONDUCTIVITY

A COMPARISON IN THERMAL-HYDRAULICS ANALYSIS OF PWR-1000 USING FIXEDAND TEMPERATURE FUNCTION OF THERMAL CONDUCTIVITY. A study to analyze theinfluence of the fuel-cladding’s thermal conductivity on the sub-channel of pressurized waterreactor 1000 (PWR-1000) using COBRA-EN computer code was conducted. The purpose ofthis research is to gain complete understanding of sub-channel thermal-hydraulic aspectsrelated to fuel performance, especially the appropriate range of thermal conductivity of UO2fuel (kf) and zircaloy-4 cladding (kc) in order to obtain an accurate sub-channel analysisrelated to its safety behavior. The research was conducted by comparing the calculation withthe combination values of the fixed kf and kc, as well as the calculation using kf and kc astemperature function. The fixed kf using in this calculation were 5.26 W/m.K, 3.85 W/m.K,3.60 W/m.K, 3.18 W/m.K, 2.90 W/m.K, 2.53 W/m.K and 2.34 W/m.K, while the kc were13.0 W/m.K, 15.57 W/m.K, 16.75 W/m.K, 17.94 W/m.K and 18.69 W/m.K. The maximumfuel center linet emperature using kf and kca s temperature function (MATPRO) for hot sub -channel was 1717.65°C and taken as the reference in accepting the calculation result usingfixed thermal conductivity. The analysis was accepted, if the deviation between bothtemperature was in the range of -10% to 10%. This analysis results for hot sub-chan nelwas accepted for the calculation using value of kf in the range of 3.18 - 2.90 W/m.K for allall variation value of kcW. hile the calculation using value of koff 2.53 W/m.K was accepet dfor value of kc in the range of 16.76 - 18.69 W/m.K

NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR)

This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR). The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers), heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s). The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use2×2radial nodes per assembly,1×18axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900°C. (2) Maximum cladding stress in circumferential direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be less than 1.0. Unfortunately some criteria could not be satisfied, such as the maximum fuel centerline temperature and maximum cladding stress. Therefore the influence of average linear heat generation rate, Pu enrichment and compressive stress to yield strength ratio are analyzed. Finally the improved fuel rod design of the new core is proposed.

Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts, one of which is supercritical water-cooled reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and light-water, graphite-moderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter water-cooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel-channel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.

Thermal Aspects of Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors

Supercritical water-cooled nuclear reactors (SCWRs) are a Generation IV reactor concept. SCWRs will use a light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). One reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is to increase the thermal efficiency. The thermal efficiency of existing NPPs is between 30% and 35% compared with 45% and 50% of supercritical water (SCW) NPPs. Another benefit of SCWRs is the use of a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing a pressure tube (PT) type SCWR. This concept refers to a 1200-MWel PT-type reactor. Coolant operating parameters are as follows: a pressure of 25 MPa, a channel inlet temperature of 350°C, and an outlet temperature of 625°C. The sheath material and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for the sheath is a zirconium alloy and the fuel is an enriched uranium dioxide (UO2). The sheath-temperature design limit is 850°C, and the industry accepted limit for the fuel centerline temperature is 1850°C. Previous studies have shown that the maximum fuel centerline temperature of a UO2 pellet might exceed this industry accepted limit at SCWR conditions. Therefore, alternative fuels with higher thermal conductivities need to be investigated for SCWR use. Uranium carbide (UC), uranium nitride (UN), and uranium dicarbide (UC2) are excellent fuel choices as they all have higher thermal conductivities compared with conventional nuclear fuels such as UO2, mixed oxides (MOX), and thoria (ThO2). Inconel-600 has been selected as the sheath material due its high corrosion resistance and high yield strength in aggressive supercritical water (SCW) at high-temperatures. This paper presents the thermalhydraulics calculations of a generic PT-type SCWR fuel channel with a 43-element Inconel-600 bundle with UC and UC2 fuels. The bulk-fluid, sheath and fuel centerline temperature profiles, together with a heat transfer coefficient profile, were calculated for a generic PT-type SCWR fuel-bundle string. Fuel bundles with UC and UC2 fuels with various axial heat flux profiles (AHFPs) are acceptable since they do not exceed the sheath-temperature design limit of 850°C, and the industry accepted limit for the fuel centerline temperature of 1850°C. The most desirable case in terms of the lowest fuel centerline temperature is the UC fuel with the upstream-skewed cosine AHFP. In this case, the fuel centerline temperature does not exceed even the sheath-temperature design limit of 850°C.

Analysis of fuel rod behavior under normal operating conditions in Super Fast Reactor

A Super Fast Reactor is a pressure-vessel type, fast spectrum supercritical water-cooled reactor (SCWR) that is presently researched in a Japanese project. A preliminary core has been designed with 1.59E+06 W/m 3 of power density [1]. In order to ensure the fuel rod integrity, the fuel rod behaviors under the normal operating conditions are analyzed using FEMAXI-6 code. Three types of the limiting fuel rods, with the maximum cladding surface temperature (MCST), maximum power peak (MPP) and maximum discharge burnup (MDB), are chosen to cover all the fuel rods in the core. The power histories of these fuel rods are taken from the neutronics calculation results in the core design. The available design range of the fuel rod design parameters, such as the initial gas plenum pressure, gas plenum length, grain size and pellet-cladding gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumstance direction should be less than 100 MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Compressive stress to yield strength ratio should be less than 0.2. (5) Cumulative damage fraction (CDF) on the cladding should be less than 1.0. Finally the improved fuel rod design is proposed.