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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents

The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms.The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. In conclusion, the failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO 2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Report on Design and Failure Limits of SiC/SiC and FeCrAl ATF Cladding Concepts under RIA

Pellet-cladding mechanical interaction (PCMI) is a possible failure scenario during a postulated reactivity-initiated accident (RIA). This report summarizes fiscal year 2018 research activities that were undertaken to evaluate the PCMI-like hoop strain–driven mechanical response of accident tolerant fuel cladding candidates of silicon carbide fiber/silicon carbide matrix composite (SiC/SiC) and iron–chromium–aluminum (FeCrAl) tube specimens. To achieve various RIA-like conditions, the modified-burst test device developed during fiscal years 2016 and 2017 was used. A heating capability was also added to the system for high-temperature testing of FeCrAl samples.The first section of this report assesses the accuracy of the two-dimensional digital image correlation calculated strain values for tube samples. The speckle patterns are computer generated in a two-dimensional plane, and the generated images were wrapped into a cylindrical surface for the validation using digital image correlation technique on curved geometries. Subsequent sections address the failure behavior of SiC/SiC composites and long-term heat treated and as-drawn FeCrAl alloy tubes at various strain rates relevant to the postulated RIA events. In the last part, the effect of possible design parameters of failure strain and the strain rate are discussed for SiC/SiC composites and FeCrAl alloys during the PCMI phase of RIA.Future work is planned to develop and design safety tests for accident tolerant fuel cladding candidates and the advanced fuel forms with respect to their transient responses. A broader perspective on the mechanical behavior of the unirradiated and irradiated accident tolerant fuel cladding candidates will be developed using the modified-burst test and other mechanical tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Progress Report on Modified Burst Testing and Alternative Test Methodologies

Performance of accident-tolerant fuel candidates must be assessed during the pellet-cladding mechanical interaction (PCMI) phase of a design-basis reactivity-initiated accident (RIA). Therefore, in this report, we describe Fiscal Year (FY) 2019 research activities on the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions comparable to the PCMI. The first section of this report provides modified-burst test (MBT) results of unirradiated FeCrAl tubes at various pressurization-rates and an RIA-relevant temperature (275°C). The failure strain of FeCrAl was observed and calculated using a high-speed digital-image correlation (DIC) technique developed recently under the Department of Energy (DOE)’s Advanced Fuel Campaign (AFC). To overcome the complexities of performing MBT in a hot-cell environment on irradiated samples, a “plane-strain tension” (PST) test was proposed as an alternative methodology for PCMI testing. The relevance of the PST test as an alternative to MBT for irradiated specimens was explicated using the data available in the literature. The design of hot-cell compatible PST fixtures and a specific holder for PST hot-cell specimen machining were developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Plane-Strain Tension Testing of ATF FeCrAl Cladding (Progress Report)

Performance of advanced accident-tolerant fuel (ATF) claddings must be assessed under all relevant scenarios, including normal and off-normal operating conditions, as well as during a reactivity-initiated accident (RIA). During RIA, the fuel pellet may expand and cause significant loads on cladding, a condition known as pellet cladding mechanical interaction (PCMI), which may even lead to cladding failure. This report describes the fiscal year (FY) 2020 research activities and results from the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions expected during RIA/PCMI events. To assess material performance under PCMI, a high strain rate burst test, often termed the modified burst test (MBT), is commonly employed. However, execution of such a test in a hot cell requires significant research. To overcome the complexities of performing MBTs in a hot-cell environment on irradiated samples, a plane-strain tension (PST) test was proposed by several authors as an alternative approach for PCMI testing. A new fixture was developed so that PST specimens could be tested using a more accommodating design for hot-cell manipulators. This report presents the PST fixture design, evaluates the possibility of modifying the specimen geometry, and discusses experimental results with unirradiated Al 6061 and FeCrAl tubes at various lengths. Furthermore, this report discusses the relevance and compatibility of the PST results to RIA scenarios and compares the PST and MBT results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

SCANAIR-BISON BENCHMARK ON CIP0-1 RIA TEST

In the frame of their research programs on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) and the Idaho National Laboratory (INL) have developed respectively SCANAIR and BISON codes to describe the thermo-mechanical behavior of irradiated fuel rods during Reactivity Initiated Accidents (RIA) in LWR. A RIA, characterized by a very rapid increase of reactivity and power in some rods of the reactor, can be schematically represented in two main phases. First, the energy deposition leads to a rapid rise of the fuel temperature which induces thermal swelling of the fuel pellets. During this phase, the Pellet Clad Mechanical Interaction (PCMI) leads to clad deformation and potentially to failure depending on fuel enthalpy increase and on the level of clad embrittlement. After the PCMI phase, the increase of clad temperature can lead to the boiling crisis of the water surrounding the rod. During the film boiling phase, the clad to coolant heat transfer becomes very low and the clad can reach high temperature (>700°C). Depending on the internal gas pressure, the ductile clad can undergo large deformation and possible failure. The RIA codes have to predict properly these two phases of the transient in order to be used, in particular, to assess current fuel safety criteria but also to be able to predict the behavior of upcoming fuel (such as accident tolerant fuel). In this paper, codes computation comparison against experimental data obtained on CIP0-1 test is done for the first phase of the transient, the PCMI phase, in order to point out the strengths and weaknesses of the codes. The CIP0-1 experiment was performed with UO2 high burnup fuel in the CABRI sodium loop facility, at 280°C and low pressure (~3 bars) without boiling crisis and no failure of the fuel rod. The initial state of the rod calculated by irradiation codes are compared with measurements before the transient. Then global rod thermal behavior during the CIP0-1 test is assessed with the sodium coolant temperature measurements. The rod mechanical behavior is analyzed with the fuel and clad elongation and the clad residual hoop strain. Finally comparison of the fission gas behavior between codes is performed and compared to the measurements of the fission gas released in the experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

An improved model for pellet-clad interaction for FRAPCON

The widely used fuel performance code FRAPCON-3 currently makes use of the rigid pellet model for Pellet-Cladding Mechanical Interactions (PCMI). Due to the potential of PCMI to limit fuel lifetime at higher burnups, the FRAPCON Radial-Axial Soft Pellet (FRASP) model is being developed for implementation into an improved version of FRAPCON, the FRAPCON-EP for Enhanced Performance. FRASP solves for the radial, hoop, and axial components of stress and strain for the fuel and cladding, including both elastic and inelastic behavior, in a quasi-static approach. Under PCMI conditions, FRASP couples the axial and radial states of the fuel and cladding via interfacial pressure and friction. Unlike the hard pellet model employed in FRACAS-I, the mechanical model used by FRAPCON-3, FRASP does allow for the stress-induced deformation of the fuel pellet. It also accounts for the radial distribution of stress and strain components through the cladding thickness. FRASP has been implemented into FRAPCON-3 and preliminarily validated against FRAPCON-3.3 for one PWR case and one BWR case with good results. The predicted fuel and cladding strains only slightly different with the new model, generally reducing the dimensional changes of the fuel pellets. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions*

Here, a reactivity-initiated accident (RIA) is a postulated design basis accident in light water reactors (LWRs) in which a rapid reactivity insertion induces a fission rate increase and a fuel pellet temperature rise. During RIA, the fuel pellet thermally expands and may cause pellet-cladding mechanical interaction (PCMI). Separate effects PCMI tests were performed on C26M FeCrAl cladding tube samples, introducing biaxial stress via a well understood modified burst test (MBT) system. A high-speed camera in the MBT system captured the projections, covering a 360° view of the cladding deformation and enabling a digital image correlation (DIC) method to quantify the surface strains with high fidelity. Representative hot zero-power RIA mechanical loading conditions were applied to the sample, and the test duration ranged from 20 to 500 ms at an average temperature of 573 K. BISON finite element–based fuel performance code modeling was performed against the high-fidelity DIC data produced from the MBTs. Validation calculations were conducted with 2D models and systematically compared with test data of the burst time, burst pressure, burst hoop strains, and hoop strain rates. Based on the behaviors from the separate effects test, BISON calculations satisfactorily predicted the cladding deformation behaviors. Sensitivity analysis was conducted to identify highly influential mechanical properties responsible for the cladding failure behavior during the MBT experiments. The results highlight the significance of the cladding's mechanical strength in governing cladding strain, followed by the significance of the mechanical interactions between the pellet and the cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO 2 ) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO 2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO 2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

RIA simulation tests using driver tube for ATF cladding

Pellet-cladding mechanical interaction (PCMI) is a potential failure mechanism for accident-tolerant fuel (ATF) cladding candidates during a reactivity-initiated accident (RIA). This report summarizes Fiscal Year (FY) 2017 research activities that were undertaken to evaluate the PCMI-like hoop-strain-driven mechanical response of ATF cladding candidates. To achieve various RIA-like conditions, a modified-burst test (MBT) device was developed to produce different mechanical pulses. The calibration of the MBT instrument was accomplished by performing mechanical tests on unirradiated Generation-I iron-chromium-aluminum (FeCrAl) alloy samples. Shakedown tests were also conducted in both FY 2016 and FY 2017 using unirradiated hydrided ZIRLO™ tube samples. This milestone report focuses on testing of ATF materials, but the benchmark tests with hydrided ZIRLO™ tube samples are documented in a recent journal article.a For the calibration and benchmark tests, the hoop strain was monitored using strain gauges attached to the sample surface in the hoop direction. A novel digital image correlation (DIC) system composed of a single high-speed camera and an array of six mirrors was developed for the MBT instrument to better resolve the failure behavior of samples and to provide useful data for validation of high-fidelity modeling and simulation tools. The DIC system enable a 360° view of a sample’s outer surface. This feature was added to the instrument to determine the precise failure location on a sample’s surface for strain predictions. The DIC system was tested on several silicon carbide fiber/silicon carbide matrix (SiC/SiC) composite tube samples at various pressurization rates of the driver tube (which correspond to the strain rates for the samples). The hoop strains for various loading conditions were determined for the SiC/SiC composite tube samples. Future work is planned to enhance understanding of the failure behavior of the ATF cladding candidates of age-hardened FeCrAl alloys and SiC/SiC composites in detail during RIA conditions informed by the computational studies performed under the US Department of Energy Office of Nuclear Energy Advanced Fuels Campaign. The testing instrument and the new DIC system will be further developed to reach different stress-state conditions and to perform tests at elevated temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

BISON Contact Algorithm Improvements in Support of Pellet Cladding Mechanical Interaction Modeling

The thermal and mechanical behavior of the gap between the fuel and cladding plays an extremely important role in determining the response of light water reactor (LWR) fuel under irradiation. This is especially true during situations when pellet-cladding mechanical interactions (PCMI) have a significant effect on the behavior of the fuel/cladding system. The BISON code, which is being used as the fuel performance simulation tool for the CASL program, employs contact detection and enforcement algorithms to realistically simulate the thermal and mechanical interactions across the fuel/cladding interaction in LWR fuel. These contact algorithms permit accurate representation of these interactions even when there is significant relative movement between the two surfaces, which is typically the case in many regions of an LWR fuel rod. This report summarizes work done during the 2017 fiscal year to further improve the accuracy and robustness of the contact enforcement algorithms in BISON. This work will benefit all BISON simulations, but will particularly benefit simulations scenarios where PCMI is of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre cracks are needed for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Post-Irradiation Examinations of the ATF Experiments - 2020 Status

This report contains the results of post-irradiation examination (PIE) of both Accident Tolerant Fuels-1 (ATF-1) rodlets and ATF-2 rodlets irradiated in the Advanced Test Reactor (ATR). The experiment is part of the U.S. Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) program’s Advanced Fuels Campaign (AFC). The ATF-1 PIE focused on the continuation of the analyses of five rodlets, one containing U 3 Si 2 fuel, two containing UN-U 3 Si 5 composites, and two with UN-U 3 Si 2 composites. The non-destructive examinations of the U 3 Si 2 rodlet showed results comparable to the previously investigated rodlets. The profilometry of the UN-U 3 Si 2 composites highlighted permanent deformation of the cladding at the pellet-pellet interface, suggesting that hourglassing of the composite pellet occurred during irradiation, causing increased localized stresses on the Zirlo cladding tube due to Pellet-Cladding Mechanical Interaction (PCMI). The UN-U 3 Si 5 rodlets, which were contained in Kanthal cladding, did not show enhanced cladding deformation at the pellet interfaces; however, it is to be noted that the initial gap of these two rodlets was larger than typical light water reactor (LWR) gaps. The larger gap might have been sufficient to accommodate the fuel swelling without having contact between the fuel and cladding, explaining the absence of PCMI. Regarding the fission products, no axial redistribution of the major gamma emitters was observed in any of the composite rodlets. Fission gas measurements on the UN-U 3 Si 5 rodlets also has been performed. The estimated fission gas release, based on the calculated fission product inventory, remains limited, with a higher fission gas release for the rodlet that experienced a higher power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

High-burnup Experiments in Reactivity Initiated Accidents (HERA)

High-burnup Experiments in Reactivity-initiated Accidents (HERA) is a joint experimental program (JEEP) operating within the United States (U.S.) Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments (FIDES). HERA is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under reactivity-initiated accidents (RIA). In-pile RIA experiments have been performed on high-burnup fuels (e.g., above 60 gigawatt days per metric ton of uranium [GWd/MTU]) in the CABRI reactor at the Cadarache site in southern France, and the Nuclear Safety Research Reactor (NSRR) in Japan. However, most of these experiments have taken place with heavily corroded Zircaloy claddings with pulse widths that are narrower (e.g., 5 ms – 30 ms full-width-half-max [FWHM]) than what would be likely in a commercial LWR (e.g., 30 ms – 80 ms FWHM). A few tests were performed in the CABRI facility with FWHM up to 76 ms, though only one “blistered” rod experienced cladding failure. Heavy waterside corrosion and narrow pulse widths are both known to increase the vulnerability of LWR fuel to pellet cladding mechanical interaction (PCMI). The HERA proposal is designed to: (1) quantify the impact of pulse width on fuel performance, offering new insight into the applicability of existing data; (2) generate new data on high burnup fuel under pulse conditions prototypic of LWRs; (3) quantify the additional margin provided by modern cladding alloys to PCMI failure limits; and (4) offer improved data for modelers using specially designed tests that eliminate key uncertainties in high-burnup fuel tests.

(HERA)↗
DOE OSTI2014

Verification of the BISON fuel performance code

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Labo- ratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze 1D spherical, 2D axisymmetric, or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON’s unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

Bison Fuel Performance Modeling on RIA in a PWR

This report is prepared to fulfill a Level 2 milestone for CASL on “deliver and document RIA fuel performance modeling using Bison fuel performance code.” The work performed is focused on modeling the full length commercial fuel rod response under a design basis Reactivity Initiated Accident (RIA), in particular, a hypothetical Control Rod Ejection Accident (CREA) in a Pressurized Water Reactor (PWR). Fuel rods with different burnups, 0 (fresh fuel), 20, 40, and 60 GWd/tU have been modeled with a reactivity worth insertion of $1.24 at Hot Zero Power (HZP) condition. Bison failure modeling method, and a brief summary of validations on selected RIA cases from CABRI and NSRR testing facilities are also described. The work performed has demonstrated the successful application of Bison code in the modeling of thermal and mechanical responses in RIAs for PWRs. In particular, Bison code has a more mature capability in the modeling of the frictional contact between pellet and cladding of the high burnup fuels, which is critical for modeling the Pellet Clad Mechanical Interaction (PCMI). The modeling of RIA using Bison code shows fuels in PWR environment could have different responses than in the simulated RIA tests, and results indicate that a transition from PCMI to Departure from Nucleate Boiling (DNB) could happen at relatively lower enthalpies: a better heat transfer from the fuel to coolant could increase cladding temperature and ductility, but at the same time also reduces the margin to DNB, which could cause high temperature oxidation and ballooning type of failures. The precise determination of the transition into DNB may rely on a more accurate input of the power histories in the simulated RIAs from VERA, the core simulator of CASL. The limiting conditions in PWRs could be quite different than those tests with non-typical testing conditions, and the improvement of the modeling of thermal responses, such as, clad coolant heat transfer, gap heat transfer, clad thermal inertia, and radial power profiles could be important considerations of the RIAs. In modeling the fuel responses in a fast transient, a number of variables could affect the initial burnup states, and some material models, e.g., fuel mechanical models, fission gas release, and corrosion models, and parameters can affect modeling results as well. Those uncertainties and advanced modeling capabilities need to be investigated in the future study of RIAs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

An evaluation of the nuclear fuel performance code BISON

BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON's unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Analyses of SPERT-CDC test 859 by FEMAXI-7 and RANNS codes

In the current Japanese regulation concerning fuel safety of water-cooled reactors, the criterion of fuel failure due to pellet-cladding mechanical interaction (PCMI) in a burnup range of 25-40 GWd/t is determined substantially based on the result of SPERT-CDC test 859 (SPERT859). However, there is room for discussion if the result of SPERT859 test rod with a burnup of 31.8 GWd/t represents the failure behavior of the fuel rod for commercial light water reactor because it was reported in some literatures that the cladding of SPERT859 test rod corroded excessively during pre-irradiation. On the other hand, the oxide thickness on the cladding of SPERT859 test rod was not reported in the data of its post-irradiation examinations (PIEs) after pre-irradiation, and the probability of the excessive corrosion on the cladding of the test rod and the effect of such excessive corrosion on the fuel enthalpy at failure of the test rod has not yet been clarified. In this study, the oxide thickness of the cladding formed on the cladding outer surface of SPERT859 test rod and its fuel enthalpy at failure due to PCMI under this corrosion condition were analyzed by using fuel performance codes FEMAXI-7 and RANNS. From the analysis results of FEMAXI-7, it was found that the excessive corrosion of the cladding of the test rod should be assumed to reproduce the fuel outer diameter measured in PIEs after pre-irradiation. The fuel enthalpy at failure of the test rod was calculated by using RANNS, and the calculated fuel enthalpy at failure was relatively close to the value of SPERT859 test rod. These results suggested that the fuel enthalpy at failure of SPERT859 was affected by the excessive corrosion on the cladding of the test rod and was likely lower than that of the typical fuel for light water reactors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗