Zircaloy Cladding Research Articles

A new procedure for solving steady-state and transient-state nonlinear radial conduction problems of nuclear fuel rods

Half-boundary method (HBM) is extended to solve one-dimensional radial conduction problems in nuclear fuel rods. The accuracy and efficiency of HBM are first tested by solving three one-dimensional radial conduction problems in cylindrical coordinates. They are steady-state and transient-state problems with linear and nonlinear material properties. HBM shows a comparable accuracy to the analytical solutions, and a higher accuracy than finite volume method (FVM) especially with fewer elements in the discretized model. Besides, the computation time ratio of HBM to FVM decreases as the element numbers increase, and it is only 0.16% when 1000 elements are involved. In order to further validate the practical applicability of HBM, nonlinear heat conduction problems of nuclear fuel rods (NFRs) are solved by HBM. The numerical results of NFRs are carried out both in steady-state and transient-state cases. HBM has a matched result with ANSYS. In addition, the size effect of pellet, helium gap and zircaloy cladding on the highest temperature in NFRs is also studied using HBM. The pellet dimension affects the highest temperature mostly.

Potential impact of accident tolerant fuel cladding critical heat flux characteristics on the high temperature phase of reactivity initiated accidents

The present paper details scoping RELAP5-3D simulations to assess the potential impact of boiling heat transfer coefficients and critical heat flux (CHF) on peak cladding temperature (PCT) of key accident tolerant fuel (ATF) cladding types in a design basis accident (DBA). The accident event studied is a Hot Zero Power (HZP) Reactivity-Initiated Accident (RIA) in a Pressurized Water Reactor (PWR), so the present study is focused on bubble crowding CHF. Results are presented for the sensitivity of PCT and the duration time of film boiling to the key parameters in the boiling curve: nucleate, transition, and film boiling heat transfer coefficient, and CHF. This included a generic study of perturbations in boiling heat transfer characteristics for Zircaloy-4 cladding as well as a study which focused on the key candidate ATF cladding types. For Zircaloy-4 cladding, the simulation results show that both PCT and duration of film boiling are particularly sensitive to the value of CHF. We also compared the performance of FeCrAl claddings and SiC/SiC claddings with Zircaloy-4 cladding under the same RIA scenario. A 10% CHF enhancement for FeCrAl cladding can significantly reduce the PCT of FeCrAl claddings. Lower irradiated thermal conductivity of SiC/SiC claddings does cause a high temperature gradient in the radial direction.

Mechanical analysis of surface-coated zircaloy cladding

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

Multispectral pyrometry for surface temperature measurement of oxidized Zircaloy claddings

Non-contact temperature measurement in a nuclear reactor is still a huge challenge because of the numerous constraints to consider, such as the high temperature, the steam atmosphere, and irradiation. A device is currently developed at CEA to study the nuclear fuel claddings behavior during a Loss-of-Coolant Accident. As a first step of development, we designed and tested an optical pyrometry procedure to measure the surface temperature of nuclear fuel claddings without any contact, under air, in the temperature range 700–850°C. The temperature of Zircaloy-4 cladding samples was retrieved at various temperature levels. We used Multispectral Radiation Thermometry with the hypothesis of a constant emissivity profile in the spectral ranges 1–1.3µm and 1.45–1.6µm. To allow for comparisons, a reference temperature was provided by a thermocouple welded on the cladding surface. Because of thermal losses induced by the presence of the thermocouple, a heat transfer simulation was also performed to estimate the bias. We found a good agreement between the pyrometry measurement and the temperature reference, validating the constant emissivity profile hypothesis used in the MRT estimation. The expanded measurement uncertainty (k=2) of the temperature obtained by the pyrometry method was ±4°C, for temperatures between 700 and 850°C. Emissivity values, between 0.86 and 0.91 were obtained.

Identification of the steady-state creep behavior of Zircaloy-4 claddings under simulated Loss-Of-Coolant Accident conditions based on a coupled experimental/numerical approach

The secondary creep behavior of Zircaloy-4 (Zr-4) claddings under simulated Loss-Of-Coolant Accident (LOCA) conditions was investigated. A coupled experimental / numerical approach is developed to determine the steady-state creep constitutive law’s parameters. The test matrix targets thermal mechanical conditions for which several creep mechanisms are expected to take place (phase transition zone of Zr-4).Ballooning tests were performed on tubular specimens. A heterogeneous thermal mechanical loading was applied to the specimen. The resulting thermal and kinematics fields were measured using full-field optical measurements. Then a Finite Element model was updated to determine a constitutive Norton creep law fitting the kinematic fields measured by way of digital image correlation (DIC). The following methodology enables a significant reduction in the number of tests used for the mapping of creep mechanisms depending on stress and temperature conditions. The results of nine experiments covering stresses from 7 to 45 MPa and temperatures varying from 745 to 850 °C are detailed. In the α-phase domain, the transition from low Norton exponents (∼ 1) to higher ones (∼ 4-5) is observed at a von Mises stress of 26 MPa. The activation energies are consistent in the α-phase domain. For modeling purposes, several tests (T < 810 °C) were then correlated. Three constitutive laws, depending on the stress and creep-rate levels, were determined. Norton exponent and activation energy consistent with the superplasticity observed by Garde et al. were identified in the early stage of the two-phase domain.

ZrC COATING ON FUEL ELEMENT CLADDING ZIRCALOY-2

ZrC COATING ON FUEL ELEMENT ZIRCALOY-2 CLADDING. The intensive researchs on high discharge burn-up of Light Water Reactor (LWR) fuel element were performed due to the extension of fuel element’s utility life. One of these researches was allowing for alteration of the existing zirconium-based clad system through coating. This technique is supposed to improve the corrosion resistance of cladding without changing the dimension of fuel cladding. In current research, the ZrC film was coated on the zircaloy-2 cladding surface by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens then undergone heating process at 700oC, 900oC and 1100oC respectively in Argon gas atmosphere for 1 hour. The microstructure and crystal structure of the coated cladding were characterized by optical microscope and XRD respectively. The optical microscope showed the growth of the grains with increasing temperature. XRD examination on the specimens revealed that the ZrC crystal structure on the cladding surface occurred only at 1100oC, but it did not appear at 700oC and 900oC. It can be concluded that dipping process of specimen in colloidal graphite with subsequent heating at 1100oC provided ZrC film coated on zircaloy-2 cladding. The heating process at this temperature allowed carbon atoms to diffuse into zircaloy surface to form ZrC film.PELAPISAN ZrC PADA KELONGSONG ELEMEN BAKAR NUKLIR ZIRKALOI-2. Riset yang intensif pada elemen bakar reaktor berpendingin air dengan fraksi bakar tinggi terus dilakukan dalam rangka memperpanjang umur operasi elemen bakar. Salah satu riset tersebut berupa proses untuk mengubah kelongsong berbasis zirkonium yang ada saat ini dengan cara pelapisan. Cara ini diharapkan akan memperbaiki ketahanan korosi kelongsong tanpa mengubah dimensi kelongsong tersebut. Pada riset ini, lapisan tipis ZrC dilapiskan pada permukaan kelongsong zirkaloi-2 melalui proses pencelupan (dipping) spesimen zirkaloi-2 dalam koloid grafit pada temperatur kamar. Spesimen yang telah dicelup dipanaskan masing-masing pada temperatur 700oC, 900oC dan 1100oC dalam media gas Argon selama 1 jam. Kelongsong yang telah dilapisi dan mengalami proses pemanasan, dikarakterisasi menggunakan mikroskop optik dan uji XRD untuk mengetahui mikrostruktur dan struktur kristalnya. Pengujian dengan mikroskop optik memperlihatkan pertumbuhan butir seiring dengan naiknya temperatur pemanasan. Pengujian dengan XRD mengungkapkan bahwa struktur ZrC terbentuk pada permukaan spesimen kelongsong zirkaloi-2 yang dipanasi pada temperatur 1100oC. Struktur ZrC tidak muncul pada spesimen hasil pelapisan dengan pemanasan pada temperatur 700oC and 900oC. Berdasarkan analisis data dari riset ini dapat disimpulkan bahwa pembentukan lapisan tipis ZrC pada permukaan spesimen kelongsong zirkaloi-2 dapat dilakukan melalui proses pencelupan spesimen dalam koloid grafit yang diikuti dengan pemanasan pada temperatur 1100oC yang memungkinkan atom-atom karbon mempunyai cukup energi untuk mendifusi ke permukaan kelongsong membentuk struktur ZrC.

Study of a “hot” particle with a matrix of U-bearing metallic Zr: Clue to supercriticality during the Chernobyl nuclear accident

This paper is dedicated to the 30th anniversary of the severe nuclear accident that occurred at the Chernobyl NPP on 26 April 1986. A detailed study on a Chernobyl “hot” particle collected from contaminated soil was performed. Optical and electron microscopy, as well as quantitative x-ray microbeam analysis methods were used to determine the properties of the sample. The results show that the particle (≈240 x 165 μm) consists of a metallic Zr matrix containing 2–3 wt. % U and bearing veins of an U,Nb admixture. The metallic Zr matrix contains two phases with different amounts of O with the atomic proportions (U,Zr,Nb)0.73O0.27 and (U,Zr,Nb)0.61O0.39. The results confirm the interaction between UO2 fuel and zircaloy cladding in the reactor core. To explain the process of formation of the particle, its properties are compared to laboratory experiments. Because of the metallic nature of the particle it is concluded that it must have formed during a very high temperature (>2400∘C) process that lasted for only a very short time (few microseconds or less); otherwise the particle should have been oxidised. Such a rapid very high temperature process indicates that at least part of the reactor core could have been supercritical prior to an explosion as it was previously suggested in the literature.

Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding alloys, Cr, ZrSi, TiAlC, and TiSiC. This ATF performance analysis platform will also be used to support experimental work underway in our Integrated Research Project.

Kinetika Pembentukan Fasa Hidrida pada Zircaloy-4 Selama Proses Gaseous Hydriding dengan Pendekatan Persamaan Avrami

HYDRIDE FORMATION KINETICS ON GASEOUS HYDRIDING OF ZIRCALOY-4 USING AVRAMI EQUATION. Zirconium alloys are mainly material for fuel cladding. Mechanical properties degradation of zircaloy-4 as nuclear fuel cladding is inevitable due to its interaction with hydrogen during normal reactor operation or transient condition. Kinetics analyses of Zircaloy-4 cladding materials at operating temperatures of 500, 550 and 600 oC under hydrogen atmosphere by usingAvrami's method to simulate brittle hydride phase formation and physical observation of the cladding after hydriding have been performed in order to observe the cladding to hydrogen interaction. Kinetic analyses showed that hydriding reaction rate follows the second order with activation energy of 35 kJ/mol. The formation of e-ZrH phase, an indication of the physical degradation of the tubes, consumed time of 180 minutes at a temperature of 500 oC, 150 minutes at 550 oC, and 100 minutes at 600 oC, respectively. Degradation mechanical properties could be seen visually on the higher hydriding temperature. The cladding materials showed that the tube swelling oC curs at a temperature of 550 oC and the tube degraded into pieces oCcurs at 600 oC. The value of mechanical degradation will be determined by mechanical testing in the next research.

Cyclic Voltammetry on Zr, Sn, Fe, Cr and Co in LiCl-KCl Salts at 500°C for Electrorefining of Irradiated Zircaloy-4 Cladding

To achieve a better understanding of Zircaloy-4 electrorefining in LiCl-KCl at 500°C, redox behaviors of Zr, Sn, Cr, Fe and Co, were examined by cyclic voltammetry using tungsten working and counter electrode. Zr metal can be produced from Zr4+ through two step reduction and ZrCl is generated as an intermediate product. As ZrCl4 concentration decreases, ZrCl would be easily reduced into Zr metal. In addition, Zr2+ formation would be unlikely at low ZrCl4 concentration. The redox reactions of the other elements are much simpler than Zr. The reduction of Zr4+ into ZrCl was close to irreversible but the reduction reactions of the other elements are close to reversible. The diffusion coefficients and apparent reduction potentials were calculated using equations for a reversible or irreversible reaction. The diffusion coefficients are in the order of 10−9 m2/s and Zr is less reductive than other elements.

Oxygen segregation in pre-hydrided Zircaloy-4 cladding during a simulated LOCA transient

Oxygen and hydrogen distributions are key elements influencing the residual ductility of zirconium-based nuclear fuel cladding during the quench phase following a Loss Of Coolant Accident (LOCA). During the high temperature oxidation, a complex partitioning of the alloying elements is observed. A finite-difference code for solving the oxygen diffusion equations has been developed by Institut de Radioprotection et de Sûreté Nucléaire to predict the oxygen profile within the samples. The comparison between the calculations and the experimental results in the mixed α+β region shows that the oxygen diffusion is not accurately predicted by the existing modeling. This work aims at determining the key parameters controlling the average oxygen profile within the sample in the two-phase regions at 1200 °C. High temperature steam oxidation tests interrupted by water quench were performed using pre-hydrided Zircaloy-4 samples. Experimental oxygen distribution was measured by Electron Probe Micro-Analysis (EPMA). The phase distributions within the cladding thickness, was measured using image analysis to determine the radial profile of α(O) phase fraction. It is further demonstrated and experimentally checked that the α-phase fraction in these regions follows a diffusion-like radial profile. A new phase fraction modeling is then proposed in the cladding metallic part during steam oxidation. The modeling results are compared to a large set of experiments including the influence of exposure duration and hydrogen content. Another key outcome from this modeling is that oxygen average profile is straightforward derived from the proposed modeling.

Mechanical Properties of Zircaloy-4 Cladding Tube by Advanced Expansion due to Compression (A-EDC) Test

Mechanical Properties of Zircaloy-4 Cladding Tube by Advanced Expansion due to Compression (A-EDC) Test

Characterization of Zircaloy-4 after Gaseous Hydriding at the Temperature Range of 350-600&lt;sup&gt;°&lt;/sup&gt;C

The degradation of the mechanical properties of zircaloy-4 as nuclear fuel cladding is inevitable due to its interaction with hydrogen during normal reactor operation. This experiment observed the occurence of hydride phases after gaseous hydriding with hydrogen at elevated temperature, and their effects were evaluated based on the material's microstructure and mechanical properties. The zircaloy-4 cladding materials were annealed for two hours. It was followed by hydriding with the pressure from 100 up to 1200 mbar at various temperatures of 350°C, 500°C,550°C and 600°C. The pressure-composition-isotherm (PCI) diagram showed that the hydrogen absorbed by specimen namely 0.18% wt, 0.38% wt, 0.44% wt and 0.74% wt at 350°C, 500°C, 550°C and 600°C, respectively. This result confirmed that the hydrogen content were 10, 128, 250, and 1357 ppm at 350°C,500°C,550°C, and 600°C, respectively, as measured by an ONH analyzer. X-ray diffraction showed that δ-hydride peaks were very weak based on fitting with the hydride database. The results from optical microscope and scanning electron microscope confirmed the presence of hydrides at the specimens, identified by the growth of needle-like structure at those temperatures. Received: 5 January 2016; Revised: 15 June 2016; Accepted: 20 August 2016

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

The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors

Advanced cladding materials with potentially enhanced accident tolerance will yield different light-water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to cladding material properties, reactor physics, and thermal hydraulics characteristics. Differences in reactor physics are driven by the fundamental properties (e.g., neutron absorption cross section in iron for an iron-based cladding) and also by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption).This paper describes three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative pressurized water reactor with both iron-chromium-aluminum (FeCrAl) and silicon-carbide fiber silicon carbide ceramic matrix composite (SiC/SiC) materials. This study shows similar RIA neutronic behavior for SiC/SiC cladding configurations versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition but with shorter pulses of higher magnitude. This is due to the shorter neutron generation time of the core models based on FeCrAl cladding. The FeCrAl-based cases exhibit a more rapid fuel thermal expansion rate than other cases, and the resultant pellet-cladding interaction may occur more rapidly. The conclusions in this paper are based on a limited set of simulated super prompt RIA transients.

Optical pyrometry measurement on oxidized Zircaloy-4 cladding

In order to improve the safety of nuclear power plant, loss-of-coolant accident experiments are implemented in research reactor. In this framework, we develop an optical pyrometry device to measure surface temperature (700-1200°C) of Zircaloy cladding without contact. The whole set-up of the simplified device (under air, without radiation) and the measurement procedure including data treatment based on bichromatic pyrometry are presented, as well as results for various temperature levels. Temperature retrieval based on the hypothesis of emissivity ratio equal to a constant, is scanned over a large wavelength range. A rather constant surface temperature is obtained on the spectral range of measurement, confirming the relevancy of emissivity hypothesis. Differences between this non-contact temperature measurement and a complementary thermocouple temperature measurement are also discussed.

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).