Fuel Analysis Code Research Articles

Areas of improvement in FRAPCON and FRAPTRAN codes for applications to fuels with cladding material change

Models and material correlations of the fuel analysis codes (FRAPCON and FRAPTRAN), which need to be modified for applying to a fuel with newly developed HANA-6 cladding are identified in the present study. The approach considering design bases of the fuel system aligned with FRAPCON and FRAPTRAN applicability to each design basis was employed and total 17 models and material correlations were identified as needing to be changed in a strict sense. Then they are re-evaluated by an expert panel based on available experimental data or sensitivity study results. This re-evaluation process leads to the final identification of 10 models and material correlations which should be corrected for effective HANA-6 cladding application. Considering diverse accident tolerant fuels employing different kinds of cladding materials are under active development now, areas of improvement of FRAPCON and FRAPTRAN codes identified for HANA-6 cladding can be used as primary points to be confirmed when developing or modifying any fuel analysis codes for accident tolerant fuels applications.

Upgradation of fission gas release model in the nuclear fuel analysis code FuSPAC

The evaluation of fuel performances in the whole life cycle is important to the fuel integrity. In our previous work, a fuel performance analysis code FuSPAC was developed and validated against the in-pile test data and FRAPCON code. In the current work, this code was upgraded by replacing the modified Forsberg-Massih for fission gas release (FGR) with a mechanism model. The microscopic behaviors related to the FGR, including the gas atom diffusion, trapping, re-solution and bubble nucleation inside the granular particles, the vacancy diffusion at the inter-granular boundaries, the grain growth and the grain recrystallization were modeled. The upgraded FuSPAC code was validated against the in-pile test data for FGR and internal void volume. The sensitivity analysis on the sub-models shown that the intra-granular fission gas diffusion, the re-solution, and the inter-granular vacancy diffusion had significant effects on the FGR. The effects from rod power can be well reflected by the upgraded code compared to the original code.

Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels

The engineering application of the fuel rod system consisting of traditional UO2 pellets and SiC accident tolerant fuel (ATF) cladding is challenged by high fuel operating temperatures, large fission gas release rates and high risks of brittle failure of the cladding. In this study, a preliminary research was carried out for performance enhancement of the fuel rods consisting of SiC cladding by replacing UO2 with high thermal conductivity fuels, including UO2/BeO, U3Si2 and UN/U3Si2. Thermo-mechanical modeling was conducted with the consideration of burnup effects for the three fuels with high thermal conductivity. An upgrade of the nuclear fuel analysis code FRAPCON4.0 was conducted to extend the simulation capability by implementing the new fuel models. Thermo-mechanical behavior of the original and improved fuel rods was simulated by the upgraded FRAPCON4.0. The simulation results show that using high thermal conductivity fuels can significantly reduce the fuel temperature and fission gas release. Furthermore, the fuel gaseous swelling is greatly mitigated due to the reduced fuel temperature, which is beneficial to prevent SiC cladding from brittle failure caused by pellet-cladding mechanical interaction. By comparison, UN/U3Si2 fuel presents the most promising and desired thermo-mechanical performance among all high thermal conductivity fuels investigated in this study.

Emergency core cooling system performance criteria for Multi-Layered Silicon Carbide nuclear fuel cladding

Multi-layered Silicon Carbide (SiC) cladding has been proposed to replace the current Zircaloy (Zr) cladding in existing nuclear power plants. Due to its ceramic property, SiC may have a different failure mechanism from that of metallic Zr clad. This characteristic should be considered in designing the plant’s safety system performance, such that the fuel’s integrity is reasonably assured during accident scenarios.This paper aims at optimizing the Emergency Core Cooling System (ECCS) performance criteria in mitigating LBLOCA accidents at a Pressurized Water Reactor (PWR) using the uranium dioxide fuel and duplex or triplex SiC cladding structures. Clad failures by brittle fractures and thermal degradation were modeled in a fuel analysis code. This fuel model was coupled with the RELAP5 plant model by adapting the fuel’s mesh configuration and formulating the effective heat transfer coefficients. A model to quantify the likelihood of clad functional failure as a combination of individual layer fractures at various axial positions was proposed. Updated SiC material properties were gathered from recent reference studies. Clad reliability estimates using the proposed model were compared with the estimates obtained using other models. The results were used to relax the criterion of required actuation timing of ECCS, such that the risk of the new plant design is less or equal to that of the initial plant configuration.Triplex clad was more reliable than duplex clad and Zr, reducing the conditional core damage probability by three orders of magnitude in the reference LBLOCA scenario. Compressive stresses due to irradiation swelling played the determining role in the vulnerabilities of each clad design. The duplex clad was vulnerable to pressure-loading stresses while triplex clad was prone to fail due to thermal shocks experienced during the LOCA reflood phase. Additionally, it was concluded that the most vulnerable axial fuel rod location in the reference LBLOCA scenario was located at the highest axial power peaking location.A sensitivity study on the ECCS timing density was conducted. The results showed that ECCS actuation timing could be extended from the initial technical specification, without increasing the fuel damage likelihood. The time extension was distributed in the order of several hundreds of seconds, which was comparable with time extensions offered by FeCrAl and Cr-coated-Zr clad designs in LBLOCA scenarios. This time extension may potentially reduce the failure probabilities of EDGs and pumps, hence improving the plant safety further.

Advances in HELIOS2 nuclear data library

Abstract The ongoing development of the HELIOS2 code system at Studsvik includes periodic updates of the nuclear data library. The library expansion includes an update of the cross section data source to ENDF/B-VIIR1, a significant expansion of the burnup chains, the addition of a more complete set of gamma production data, and the development of new resonance treatment options. The goal is to provide the capability for HELIOS2 to more accurately model a wider array of reactor applications and enhance interoperability with SNF, the Studsvik spent fuel analysis code. This paper will also provide a discussion of the nuclear data library benchmarking effort and an overview of other HELIOS2 development efforts.

Thermal Conductivity Change in High Burnup MOX Fuel Pellet

High burnup MOX and UO2 test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO2 test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO2 fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.

Principle of rationalizing the criteria for abnormal transients of the Super LWR with fuel rod analyses

A detailed fuel rod design is carried out for the first time in the development of Supercritical-pressure Light Water Reactor (Super LWR). The fuel rod design is similar to that of LWR, consisting of UO 2 pellets, a gas plenum and a Stainless Steel Cladding. The principle of rationalizing the criteria for abnormal transients of the Super LWR is developed. The fuel rod integrities can be assured by preventing plastic strains on the cladding, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel analysis code is used to evaluate the fuel rod integrities in abnormal transient conditions. Detailed analyses have shown that allowable limits to the maximum fuel rod power and maximum cladding temperature can be determined to assure the fuel integrities. These limits may be useful in the plant safety analyses to confirm the fuel integrities during abnormal transients.

Evaluation of the FRANCO finite element fuel rod analysis code

Knowledge of the temperature distribution in a nuclear fuel rod is required to predict the thermal and mechanical response of the strongly temperature-dependent fuel elements. In this research, the FRANCO ( F inite Element Fuel R od AN alysis CO de) computer code was developed for use on an IBM-PC to predict thermal and mechanical fuel performance characteristics for reactor operating conditions. A cross-sectional area of a fuel rod is discretized using constant strain triangular elements. To simplify the time-dependent fuel performance analysis, this code uses quasi-steady state conditions or slow power changes for time-independent problems; thus, time-dependent cases can be represented by quasi-static “snap shots.” The development and testing of the FRANCO program for deformation and heat transfer problems is described. The fuel performance models and solution procedures are provided in detail. The accuracy of FRANCO is examined through selected benchmark problems with other nuclear industry finite difference and finite element fuel analysis codes and actual Halden experimental measurements. These benchmarks show that FRANCO performs well except for the fuel-clad contact cases when compared to results from other fuel performance codes, such as FREY and ESCORE. The FRANCO program developed in this research effort is easy to run, fast, and yields comparable results.