Materials Science › Materials Chemistry

Nuclear Materials and Properties

Works Count: 91422

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Description

This cluster of papers focuses on the development and properties of advanced materials for nuclear fuel, including accident tolerant fuels, fuel cladding, thermophysical properties, oxidation resistance, behavior under high temperature and irradiation, hydrogen embrittlement, and fission product release.

Keywords

Accident Tolerant Fuels; Fuel Cladding; Thermophysical Properties; Oxidation Resistance; Actinide Oxides; High Temperature Behavior; Density Functional Theory; Irradiation Performance; Hydrogen Embrittlement; Fission Product Release

Comprehensive Nuclear Materials

2020-01-01

R.J.M. Konings , Todd R. Allen , R.E. Stoller +1 more

Thermodynamics of nuclear materials

1963-09-01

D.A. Young

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

2015-10-23

Yukinori Yamamoto , Bruce A. Pint , Kurt A. Terrani +3 more

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Corrosion of Zirconium Alloys Used for Nuclear Fuel Cladding

2015-05-04

Arthur T. Motta , Adrien Couet , Robert J. Comstock

During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the … During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In zirconium alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup increases when electron transport decreases, requiring hydrogen ingress to close the reaction.

Multidimensional multiphysics simulation of nuclear fuel behavior

2012-01-25

R.L. Williamson , Jason Hales , Stephen Novascone +5 more

Thermophysical properties of uranium dioxide

2000-03-01

J.K. Fink

Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys

2005-01-06

B. Cox

Fuel corrosion processes under waste disposal conditions

2000-11-01

David W. Shoesmith

Metallic fast reactor fuels

1997-01-01

G.L. Hofman , L.C. Walters , T.H. Bauer

X. Quantitative measurement of preferred orientation in rolled uranium bars

1952-01-01

Glenn Harris

Synopsis The preferred orientation of uranium bars hot-rolled and cold-rolled in the α range has been determined quantitatively by x-ray diffraction using a Geiger counter. It is shown that hot … Synopsis The preferred orientation of uranium bars hot-rolled and cold-rolled in the α range has been determined quantitatively by x-ray diffraction using a Geiger counter. It is shown that hot rolling with a slight reduction causes the 010 planes to become strongly and the 110 planes weakly oriented perpendicular to the rolling direction. Further hot rolling causes the 110 planes to become strongly and the 010 planes weakly oriented perpendicular to the rolling direction. Cold rolling with light reduction causes the 010 planes to become strongly and the 130 planes weakly oriented in the planes perpendicular to the rolling direction.

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

2005-01-07

Mohamed S. El‐Genk , Jean‐Michel Tournier

Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

2013-07-01

Kurt A. Terrani , S.J. Zinkle , Lance L. Snead

Thermal annealing of fission tracks in apatite 2. A quantitative analysis

1987-03-01

G.M. Laslett , P.F. Green , Ian R. Duddy +1 more

The chemical state of the fission products in oxide fuels

1985-04-01

H. Kleykamp

Structural materials challenges for advanced reactor systems

2008-12-09

Pascal Yvon , Frank Carré

Modified embedded-atom potentials for cubic materials and impurities

1992-08-01

M. I. Baskes

In a comprehensive study, the modified embedded-atom method is extended to a variety of cubic materials and impurities. In this extension, all functions are analytic and computationally simple. The basic … In a comprehensive study, the modified embedded-atom method is extended to a variety of cubic materials and impurities. In this extension, all functions are analytic and computationally simple. The basic equations of the method are developed and applied to 26 elements: ten fcc, ten bcc, three diamond cubic, and three gaseous materials. The materials modeled include metals, semiconductors, and diatomic gases, all of which exhibit different types of bonding. Properties of these materials, including equation of state, elastic moduli, structural energies and lattice constants, simple defects, and surfaces, are calculated. The formalism for applying the method to combinations of these elements is developed and applied to the calculation of dilute heats of solution. In all cases, comparison is made to experiment or higher-level calculations when possible.

Solution and Diffusion of Hydrogen in Tungsten

1969-05-01

Regula Frauenfelder

Using mass spectroscopic and ultrahigh vacuum techniques, solution and diffusion of hydrogen in tungsten was investigated for pressures between 600 and 10−8 Torr and temperatures between 1100 and 2400 K. … Using mass spectroscopic and ultrahigh vacuum techniques, solution and diffusion of hydrogen in tungsten was investigated for pressures between 600 and 10−8 Torr and temperatures between 1100 and 2400 K. Solubility and diffusion constants are derived from degassing rates of a solid cylinder which was pre-loaded with hydrogen at ≈600 Torr. The solubility constant, S=2.9×10−1×exp (−24000/RT) Torr liter/cm3Torr1/2, and the diffusion constant, D=4.1×10−3×exp (−9000/RT) cm2/sec, are obtained, which in conjunction with literature values for the permeation constant P are consistent with the equation P=SD. Comparison to theory indicates that the solubility and diffusion constants are characteristic of interstitially dissolved hydrogen. Expressions are derived for the concentration of interstitial hydrogen as a function of pressure and temperature. Semiquantitative values for the total hydrogen concentration at low pressures are derived from H2 pressure changes which result when a sample is flashed between selected, high temperatures. Below 10−4 Torr, the total hydrogen concentration appears to be several orders of magnitude higher than the concentration of interstitial hydrogen, indicating that hydrogen is held in, and diffuses via both interstitial sites and a second kind of site of unknown nature. A semiquantitative analysis of diffusion is given for the case of the diffusing species held in, and diffusing via, two different kinds of sites.

High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments

2013-06-01

Bruce A. Pint , Kurt A. Terrani , Michael P. Brady +2 more

Voids in Irradiated Stainless Steel

1967-11-01

C. Cawthorne , E.J. Fulton

Fundamental aspects of nuclear reactor fuel elements

1976-01-01

D.R. Olander

The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and … The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

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The role of vacancies and dislocations in the nucleation and growth of gas bubbles in irradiated fissile material

1959-12-01

G. W. Greenwood , A.J.E. Foreman , D E Rimmer

Accident tolerant fuels for LWRs: A perspective

2013-12-13

S.J. Zinkle , Kurt A. Terrani , Jess C Gehin +2 more

Interstitial Atomic Diffusion Coefficients

1949-10-15

C. A. Wert , Clarence Zener

An attempt is made to interpret the temperature independent factor ${D}_{0}$ of the previously determined diffusion coefficients of interstitial solute atoms in metals. The primary uncertainty in the value of … An attempt is made to interpret the temperature independent factor ${D}_{0}$ of the previously determined diffusion coefficients of interstitial solute atoms in metals. The primary uncertainty in the value of ${D}_{0}$ given by the standard reaction rate theory resides in an entropy factor $\mathrm{exp}(\frac{\ensuremath{\Delta}S}{R})$. When cognizance is taken of an additional strain in the lattice surrounding a solute atom as it passes over a potential energy divide, and of the increase in entropy associated with an increase in lattice strain energy, one can estimate a "theoretical" range within which these entropy factors should lie. All past observations except for C and N in $\ensuremath{\alpha}\ensuremath{-}\mathrm{Fe}$ are consistent with this theoretical range. The ${D}_{0}'\mathrm{s}$ for these two systems were, therefore, redetermined by more precise measurements, and are found to be an order of magnitude higher than the original values. The associated entropy factors are consistent with the theoretical range.

On the use of SRIM for computing radiation damage exposure

2013-06-06

R.E. Stoller , M.B. Toloczko , Gary S. Was +3 more

X-ray studies of deformed metals

1959-01-01

Β. E. Warren

A review of steel corrosion by liquid lead and lead–bismuth

2009-03-20

Jinsuo Zhang

Terminal solubility and partitioning of hydrogen in the alpha phase of zirconium, Zircaloy-2 and Zircaloy-4

1967-06-01

J.J. Kearns

Materials challenges in nuclear energy

2013-01-19

S.J. Zinkle , Gary S. Was

A review of microstructure evolution in zirconium alloys during irradiation

1988-10-01

M. Griffiths

Mineralogical and chemical relationships among enstatite chondrites

1968-11-15

Klaus Keil

Fifteen of the sixteen known enstatite chondrites were studied microscopically in reflected and transmitted light, and their modal compositions were determined by point-counting techniques. Compositions of clinoenstatite, orthoenstatite, plagioclase, kamacite, … Fifteen of the sixteen known enstatite chondrites were studied microscopically in reflected and transmitted light, and their modal compositions were determined by point-counting techniques. Compositions of clinoenstatite, orthoenstatite, plagioclase, kamacite, taenite, troilite, oldhamite, daubreelite, niningerite, ferroan alabandite, and schreibersite were determined with the electron microprobe X-ray analyzer. Chemical composition, mineral occurrence, and mineral composition were found to depend on degree of recrystallization of the chondrites as judged by, for example, distinctness of chondrules and coarseness of silicates. On the basis of these parameters, three groups of enstatite chondrites can be distinguished and are referred to as type I, intermediate type, and type II. Differences between types I and II are pronounced, whereas intermediate type is transitional. The suggestion of Van Schmus and Wood that type II enstatite chondrites originated from type I by reheating is reviewed in the light of the new data. It is concluded that, although many of the chemical-mineralogical parameters of type II chondrites could be explained as being the result of reheating of type I chondrites, there are some that would require rather stringent environmental conditions during reheating. For example, lower iron and sulfur contents in type II chondrites would presumably require reheating of type I chondrites to ≥975°C, the lowest temperature at which a melt would appear in the Fe-Ni-S system of type I composition and at which physical separation of the liquid from the silicates could occur. Differences in Si/Mg ratios would require reheating to even higher temperatures and fractionation in an open system. Furthermore, observed differences in nitrogen and sinoite contents between type I and type II are difficult to explain unless reheating took place in a closed system, or under oxygen fugacities low enough to allow nitrogen to react with SiO2 and Si to form Si2N2O. An alternative model to the one by Van Schmus and Wood is discussed; it assumes that major differences in chemical and mineralogical composition between type I and type II were essentially established before or during chondrule formation and agglomeration by, for example, igneous differentiation or fractionation during condensation from a solar nebula, and that differences in texture are due either to different cooling rates of type I and type II chondrites during and after agglomeration of chondrules or to mild reheating to temperatures ≤975°C. This model does not, however, readily explain why only enstatite chondrites of type II bulk chemical composition (i.e. low Fe, S) cooled slowly or were reheated, and why chondrites of type I composition (high Fe, S) were always quenched to temperatures low enough to prevent recrystallization and were not reheated.

Hydrogen absorption and desorption properties of AB2 laves-phase pseudobinary compounds

1977-05-01

D. Shaltiel , I. Jacob , D. Davidov

On the Nature of Radiation Damage in Metals

1954-08-01

J. A. Brinkman

The nature of the permanent damage retained in metals from irradiation has been investigated in somewhat greater detail than has been done in the past. The usual assumption has been … The nature of the permanent damage retained in metals from irradiation has been investigated in somewhat greater detail than has been done in the past. The usual assumption has been that the damage in all metals consists chiefly of interstitial-vacancy pairs. The model presented in this paper reduces to this picture for the light elements but introduces a new concept in the case of damage in the heavy metals, called a displacement spike. Calculations are made from which one can estimate the relationship between the density of interstitial-vacancy pairs and the temperature of the associated thermal spike. An assumption regarding the extent to which interstitial-vacancy pairs persist throughout the duration of the thermal spike has been made, based upon these calculations. The number of interstitial-vacancy pairs predicted in the heavy elements is considerably smaller than that predicted by the former model. A mechanism is proposed by which small dislocation loops can be produced in the heavier metals by irradiation. This article is based upon studies conducted for the U. S. Atomic Energy Commission under Contract AT-11–1-GEN-8.

Structure and chemical bonding in zintl-phases containing lithium

1990-01-01

Reinhard Nesper

Structural materials for fission & fusion energy

2009-11-01

S.J. Zinkle , Jeremy T. Busby

Structural materials represent the key for containment of nuclear fuel and fission products as well as reliable and thermodynamically efficient production of electrical energy from nuclear reactors. Similarly, high-performance structural … Structural materials represent the key for containment of nuclear fuel and fission products as well as reliable and thermodynamically efficient production of electrical energy from nuclear reactors. Similarly, high-performance structural materials will be critical for the future success of proposed fusion energy reactors, which will subject the structures to unprecedented fluxes of high-energy neutrons along with intense thermomechanical stresses. Advanced materials can enable improved reactor performance via increased safety margins and design flexibility, in particular by providing increased strength, thermal creep resistance and superior corrosion and neutron radiation damage resistance. In many cases, a key strategy for designing high-performance radiation-resistant materials is based on the introduction of a high, uniform density of nanoscale particles that simultaneously provide good high temperature strength and neutron radiation damage resistance.

Development of an interatomic potential for phosphorus impurities in -iron

2004-06-26

Graeme J. Ackland , Mikhail I. Mendelev , David J. Srolovitz +2 more

We present the derivation of an interatomic potential for the iron phosphorus system based primarily on {\it ab initio} data. Transferrability in this system is extremely problematic, and the potential … We present the derivation of an interatomic potential for the iron phosphorus system based primarily on {\it ab initio} data. Transferrability in this system is extremely problematic, and the potential is intended specifically to address the problem of radiation damage and point defects in iron containing low concentrations of phosphorus atoms. Some preliminary molecular dynamics calculations show that P strongly affects point defect migration.

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Energetics and formation kinetics of helium bubbles in metals

1983-01-01

H. Trinkaus

Abstract Recent theoretical work on the energetics and formation kinetics of helium bubbles in metals is reviewed. The energetics are discussed in particular for bubbles with radii between 10 and … Abstract Recent theoretical work on the energetics and formation kinetics of helium bubbles in metals is reviewed. The energetics are discussed in particular for bubbles with radii between 10 and 1000 Å containing helium under very high pressures (nonideal gas bubbles). For this size range, the bubble formation free energy is split into three parts: a helium bulk free energy which is considered for solid and fluid helium (high-density equation of state), a bubble-matrix interface free energy for which curvature corrections are introduced, and a relaxation energy which is treated within the elastic continuum approach. The formation kinetics are considered for two extreme cases. For high helium production rates and low temperatures, helium clustering is associated with athermal processes such as metal interstitial emission and dislocation loop punching. For dislocation loop punching, an interpolation between computer simulation and continuum-theory results is discussed. For low helium production rates and high temperatures, helium clustering is connected with vacancy absorption. For this case, nucleation kinetics are briefly sketched while growth kinetics are analyzed in some detail with special reference to the effects of the high helium densities in the bubbles. In particular, the transition from gas-driven to vacancy supersaturation-driven bubble growth and its role for the evolution of bimodal bubble size distributions are analyzed. Reliable criteria to to discriminate between the two main bubble-coarsening mechanisms, bubble coalescence and Ostwald ripening, are suggested. The procedure to derive the activation energy of gas permeation from bubble coarsening, identified as Ostwald ripening is discussed.

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

2008-09-14

K.L. Murty , Indrajit Charit

Radiation-induced voids in metals

1973-01-01

R.W. Cahn

Radiation damage tolerant nanomaterials

2013-11-01

Irene J. Beyerlein , A. Caro , Michael J. Demkowicz +3 more

Designing a material from the atomic level to achieve a tailored response in extreme conditions is a grand challenge in materials research. Nanostructured metals and composites provide a path to … Designing a material from the atomic level to achieve a tailored response in extreme conditions is a grand challenge in materials research. Nanostructured metals and composites provide a path to this goal because they contain interfaces that attract, absorb and annihilate point and line defects. These interfaces recover and control defects produced in materials subjected to extremes of displacement damage, impurity implantation, stress and temperature. Controlling radiation-induced-defects via interfaces is shown to be the key factor in reducing the damage and imparting stability in certain nanomaterials under conditions where bulk materials exhibit void swelling and/or embrittlement. We review the recovery of radiation-induced point defects at free surfaces and grain boundaries and stabilization of helium bubbles at interphase boundaries and present an approach for processing bulk nanocomposites containing interfaces that are stable under irradiation.

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Electron impact phenomena

1958-04-01

L. Muldawer

Fundamentals of Radiation Materials Science

2016-07-08

Gary S. Was

Protective coatings on zirconium-based alloys as accident-tolerant fuel (ATF) claddings

2017-06-14

Chongchong Tang , Michael Stueber , Hans J. Seifert +1 more

Abstract Surface-modified zirconium (Zr)-based alloys, mainly by fabricating protective coatings, are being developed and evaluated as accident-tolerant fuel (ATF) claddings, aiming to improve fuel reliability and safety during normal operations, … Abstract Surface-modified zirconium (Zr)-based alloys, mainly by fabricating protective coatings, are being developed and evaluated as accident-tolerant fuel (ATF) claddings, aiming to improve fuel reliability and safety during normal operations, anticipated operational occurrences, and accident scenarios in water-cooled reactors. In this overview, the performance of Zr alloy claddings under normal and accident conditions is first briefly summarized. In evaluating previous studies, various coating concepts are highlighted based on coating materials, focusing on their performance in autoclave hydrothermal corrosion tests and high-temperature steam oxidation tests. The challenges for the utilization of coatings, including materials selection, deposition technology, and stability under various situations, are discussed to provide some valuable guidance to future research activities.

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Accident tolerant fuel cladding development: Promise, status, and challenges

2018-01-13

Kurt A. Terrani

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Primary radiation damage: A review of current understanding and models

2018-10-20

K. Nordlund , S.J. Zinkle , Andrea E. Sand +7 more

Scientific understanding of any kind of radiation effects starts from the primary damage, i.e. the defects that are produced right after an initial atomic displacement event initiated by a high-energy … Scientific understanding of any kind of radiation effects starts from the primary damage, i.e. the defects that are produced right after an initial atomic displacement event initiated by a high-energy particle. In this Review, we consider the extensive experimental and computer simulation studies that have been performed over the past several decades on what the nature of the primary damage is. We review both the production of crystallographic or topological defects in materials as well as radiation mixing, i.e. the process where atoms in perfect crystallographic positions exchange positions with other ones in non-defective positions. All classes of materials except biological materials are considered. We also consider the recent effort to provide alternatives to the current international standard for quantifying this energetic particle damage, the Norgett-Robinson-Torrens displacements per atom (NRT-dpa) model for metals. We present in detail new complementary displacement production estimators ("athermal recombination corrected dpa", arc-dpa) and atomic mixing ("replacements per atom", rpa) functions that extend the NRT-dpa, and discuss their advantages and limitations.

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Sputtering by Particle Bombardment II

1983-01-01

Rainer Behrish

Sputtering by Particle Bombardment I

1981-01-01

W. Jacob , J. Roth

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Ceramic steel?

1975-12-01

R. C. Garvie , R. H. J. Hannink , R. T. Pascoe

Fundamentals of Radiation Materials Science: Metals and Alloys

2007-09-15

Gary S. Was

Part I Radiation Damage -- 1 The Radiation Damage Event -- 2 The Displacement of Atoms -- 3 The Damage Cascade -- 4 Point Defect Formation and Diffusion -- 5 … Part I Radiation Damage -- 1 The Radiation Damage Event -- 2 The Displacement of Atoms -- 3 The Damage Cascade -- 4 Point Defect Formation and Diffusion -- 5 Radiation-Enhanced and Diffusion Defect Reaction Rate Theory -- Part II Physical Effects of Radiation Damage -- 6 Radiation-Induced Segregation -- 7 Dislocation Microstructure -- 8 Irradiation-Induced Voids and Bubbles -- 9 Phase Stability Under Irradiation -- 10 Unique Effects of Ion Irradiation -- 11 Simulation of Neutron Irradiation Effects with Ions -- Part III Mechanical Effects of Radiation Damage -- 12 Irradiation Hardening and Deformation -- 13 Irradiation Creep and Growth -- 14 Fracture and Embrittlement -- 15 Corrosion and Stress Corrosion Cracking Fundamentals -- 16 Effects of Irradiation on Corrosion and Environmentally Assisted Cracking -- Index. .

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Assessing the near-surface diffusion of Xe and Kr in Zirconia by time-of-flight elastic recoil detection analysis

2025-06-25

N.P. Wikstrom , Maria Giamouridou , Elina Charatsidou +4 more | Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms

Influence of thermal aging on the deuterium permeation and retention behavior in the CLAM Steel

2025-06-24

Zongming Shao , Hao Yang , Siwei Zhang | Fusion Engineering and Design

Study on the irradiation mechanical behavior of three-petal fuel rods contact based on multi-physics field coupling

2025-06-24

Weihua Cai , Shusong Qin , Binxian He +5 more | Annals of Nuclear Energy

Machine learning enhanced quantum molecular dynamics assessment of the uranium phase diagram

2025-06-24

NULL AUTHOR_ID | Physical review. B./Physical review. B

A comprehensive study on resistance brazing of bearing pad to fuel tube of pressurized heavy water reactor using amorphous filler

2025-06-23

Suman Neogy , Aniruddha Kumar , Harish Donthula +7 more | Nuclear Engineering and Design

Thermophysical Properties of a Uranium Nitride–Molybdenum-Tungsten Composite Fuel for Nuclear Thermal Propulsion

2025-06-23

Joseph Schaeperkoetter , Scarlett Widgeon Paisner , Joshua T. White +6 more | Nuclear Technology

Role of Accelerated Burnup Irradiation Testing in Support of Accelerated Fuel Qualification

2025-06-23

Andrew Nelson , Denise Adorno Lopes , Nathan Capps +1 more | Nuclear Technology

Development of heat conduction model for coated fuel particles randomly distributed in a matrix

2025-06-21

Shuoting Zhang , Jianhua Dong , Zheng Huang +2 more | Nuclear Engineering and Design

The potential of using chromium plating to reduce ingress of hydrogen isotopes in CANDU reactor rolled joints

2025-06-20

G.A. McRae , C.E. Coleman , Mui Hoon Nai +1 more | Journal of Pressure Vessel Technology

Abstract Some recent measurements of hydrogen isotope concentrations in zirconium pressure tubes at the rolled joints in CANDU reactors have been surprisingly high reaching values above the current license limit … Abstract Some recent measurements of hydrogen isotope concentrations in zirconium pressure tubes at the rolled joints in CANDU reactors have been surprisingly high reaching values above the current license limit of 120 mg H/kg Zr. These high concentrations have been attributed, in part, to corrosion and concomitant ingress of hydrogen isotopes into the pressure tube. The rolled joint consists of a stainless-steel end fitting into which a zirconium tube is rolled to form a leak-free joint. Chromium plating of the bore of the end fittings is being considered to lessen hydrogen isotope concentrations in the zirconium by reducing galvanic corrosion and by reducing direct contact between the end-fitting and the pressure tube. The experiments described in this study were all done dry, without electrolyte present. The benefit of Cr plating under dry conditions is the same as when an electrolyte is present; hence, galvanic corrosion is not contributing to the high hydrogen isotope concentrations in the rolled joints because corrosion requires an electrolyte. Instead, ingress of hydrogen isotopes happens when hydrogen gas can move between metals when their surface oxides fail under anoxic conditions. The high concentrations are caused by permeation, not corrosion. Chromium plating of the stainless-steel end fitting is not a robust solution and does not provide a reliable barrier to hydrogen movement into the zirconium.

An Update on the Tender X-Ray Spectroscopy of the Actinides

2025-06-20

James Tobin | Synchrotron Radiation News

An Assessment of Replacing Aluminum Tubes Hosting Nuclear Fuels with Stainless Steel in a Subcritical Nuclear Reactor

2025-06-19

Diego Medina-Castro , Héctor René Vega-Carrillo , Antonio Baltazar-Raigosa +3 more | Energies

This computational study using MCNP5 evaluated the feasibility of replacing 6061-T6 aluminum with 316L stainless steel (SS-316L) for the tubes hosting the uranium slugs in the subcritical nuclear reactor Nuclear … This computational study using MCNP5 evaluated the feasibility of replacing 6061-T6 aluminum with 316L stainless steel (SS-316L) for the tubes hosting the uranium slugs in the subcritical nuclear reactor Nuclear Chicago model 9000, thereby contributing to its preservation as a key resource for nuclear research and education in Mexico. Simulations and dosimetric analyses (ICRP/ICRU) confirmed subcriticality in both configurations. Notably, SS-316L demonstrated an effective attenuation of peripheral gamma radiation and a reduction in the ambient neutron dose, indicating a considerable improvement in radiological safety. Although a reduction in thermal and epithermal neutron fluence was observed, the similarity in the gamma spectrum suggests no significant alteration for gamma spectroscopic experiments. In conclusion, SS-316L presents a promising alternative that enhances radiological safety and reactor longevity, making it a worthy consideration as a replacement material. Further experimental investigation is recommended to assess material activation and the gamma dose in the vicinity of the fuel.

Multi-scale mechanical model for U-Si/Al fuels under irradiation based on evolution of bubbles

2025-06-18

Song Yue , Meng Lv , Heng Xie | Nuclear Engineering and Design

Energetic stability of small defect clusters in α-zirconium by first-principles calculation

2025-06-18

Wenlong Liu , Chen Zhang , Yizhuo Zhang +6 more | Materials Letters

Investigating the Role of Chromium-Coated Clad During Accident Progression for an SMR Using MELCOR

2025-06-18

Mohammad Amer Allaf , Grace Ejnik , Emile Zaccomer +3 more | Nuclear Technology

Calculating the Bowing-Related Reactivity Worth Using PERSENT

2025-06-18

M. A. Smith | Nuclear Technology

CFD analysis on the safety performance of CSNS solid target under a postulated large loss of coolant accident

2025-06-18

Zheng Yang , Songbai Cheng , Youlian Lu +5 more | Annals of Nuclear Energy

Disposal of large quantities of plutonium and thorium utilization in VVER-1200 reactors using americium-stabilized MOX fuel

2025-06-18

Ayhan Kara , Emil Mammadzada | Annals of Nuclear Energy

Microstructure and Mechanical Properties of Tungsten Zircaloy-4 Diffusion Welding Interface

2025-06-16

Shaohong Wei , Yan Li , Ruiqiang Zhang +3 more | Materials

The tungsten target block is widely used as a target material in spallation neutron sources. However, due to the poor corrosion resistance of tungsten, a corrosion-resistant metal layer needs to … The tungsten target block is widely used as a target material in spallation neutron sources. However, due to the poor corrosion resistance of tungsten, a corrosion-resistant metal layer needs to be coated on the surface. In this study, Zircaloy-4 coating on tungsten was prepared by hot isostatic pressure diffusion welding in the temperature range of 900 °C to 1400 °C. The microstructure and mechanical properties of the zirconium-tungsten interface were studied. The results show that a clear intermediate diffusion layer was formed at the interfaces, and no obvious defects were found. As the HIP temperature increased from 900 °C to 1400 °C, the thickness of the diffusion layer gradually increased from 0.28 μm to 10.74 μm. Composition and phase structure analysis of the intermediate diffusion layer showed that the main phase of the diffusion layer is ZrW2. The nanoindentation hardness results near the interface showed that the hardness of the ZrW2 diffusion layer was significantly higher than that of W and the zirconium alloy, reaching around 17.96 GPa. As the HIP temperature increased, the bonding strength between Zry-4 and W matrix first increased and then decreased, with the highest bonding strength of 83.9 MPa when the HIP temperature was 1000 °C.

Evaluating Radiation Characteristics of Cr-Coated Silicon Carbide and Zircaloy Claddings in a PWR Fuel Assembly: A Simulation Study Employing GEANT4 and EpiXS

2025-06-16

M.M. Mostafa , Abdus Sattar Mollah | International Journal of Chemical Engineering and Materials

Nuclear fuel cladding materials act as protective barriers for UO₂ pellets, providing essential corrosion resistance and maintaining low neutron and gamma radiation absorption cross-sections. To enhance the accident-tolerant fuel (ATF) … Nuclear fuel cladding materials act as protective barriers for UO₂ pellets, providing essential corrosion resistance and maintaining low neutron and gamma radiation absorption cross-sections. To enhance the accident-tolerant fuel (ATF) framework and strengthen nuclear fuel cladding protection, this study investigates chromium-coated Zirconium alloy (E110) and Silicon Carbide (SiC) as cladding materials in a PWR reactor fuel assembly. The key concern addressed is the degradation of the irradiated inner and outer cladding surface, driven by fission fragments produced during nuclear reactions, steam with high temperature, and the formation of hydrides. SiC demonstrates superior resistance to chemical attacks compared to Zircaloy, oxidizes less aggressively at high temperatures, has a higher yield strength, and shows a significantly lower creep rate and neutron capture cross-section. However, the application of a chromium coating can influence the radiation response of both materials. The study aims to determine the suitable nuclear fuel cladding between Zircaloy and Silicon Carbide with Cr coating. The performance of Zircaloy (E110) and SiC cladding materials with thin chromium coatings is modeled and assessed in this work using GEANT4 simulation software. The outcomes are verified and compared to the EpiXS software dataset to guarantee precision and dependability. For modeling radiation interactions and material deterioration, GEANT4 provides a sizable dataset. These discoveries aid in creating stronger cladding materials, which raise nuclear reactor efficiency and safety. The work also shows how GEANT4 can precisely evaluate the radiation characteristics of materials exposed to gamma and neutron radiation, providing vital data for improving cladding performance in radiation environments and raising nuclear safety requirements.

Deuterium retention in cold spray tantalum coatings vs. polycrystalline tungsten and tantalum

2025-06-16

Mykola Ialovega , M. Navarro , R. Bisson +7 more | Nuclear Fusion

Abstract Enhanced deuterium retention in tantalum (Ta) cold spray coatings, compared to reference polycrystalline tantalum and tungsten materials, has been evaluated using the thermal desorption spectrometry technique. Tantalum coatings, deposited … Abstract Enhanced deuterium retention in tantalum (Ta) cold spray coatings, compared to reference polycrystalline tantalum and tungsten materials, has been evaluated using the thermal desorption spectrometry technique. Tantalum coatings, deposited via cold spray technology on 316L stainless steel substrates, are proposed as plasma-facing material surfaces with hydrogen gettering functionality for advanced fusion concepts. The materials were exposed to 95 eV D ions at a flux of 1.6-3.5×10 21 D m -2 s -1 . Retention was measured as a function of incident ion fluence and surface temperature. The results highlight an increased deuterium inventory in Ta cold spray coatings by a factor of 3.5 compared to polycrystalline tantalum and by two orders of magnitude compared to polycrystalline tungsten. A tendency for retention saturation in tantalum is observed at a fluence above 1×10 24 D m -2 . While deuterium retention gradually decreases with increasing surface temperature from 400 K to 925 K for polycrystalline tungsten, it remains constant for polycrystalline tantalum. In contrast, retention in Ta coatings significantly decreases when the surface temperature exceeds 750 K. The microstructure of the cold spray Ta coatings plays a crucial role in the dynamics of deuterium trapping and release. Tantalum also exhibits a superior resistance to blister formation compared to tungsten when subjected to a high dose of deuterium.

High-temperature oxidation failure in reactivity-initiated accidents: An evaluation of failure criteria based on oxygen concentration from the previous NSRR experiments

2025-06-16

Vu Nhut Luu , Yoshinori Taniguchi , Yutaka Udagawa +1 more | Nuclear Engineering and Design

Influence of steam flow rate on oxidation kinetics of silicon carbide at 1400–1600 °C

2025-06-14

Hai Vu Pham , Masaki Kurata , Yuji Nagae +2 more | Corrosion Science

Impact of the Uncertainties Associated with Core Degradation on the State of the Corium at Vessel Rupture

2025-06-13

L. Carénini , F. Fichot | Nuclear Technology

Intra-granular fission gas bubbles growth in crystalline U 3 Si 2: rate theory modeling

2025-06-13

Cong Ma , Y.J. Pei , Tianyuan Xin +6 more | Chinese Physics B

Abstract A model of intra-grain fission gas bubble growth in U 3 Si 2 coupled with defect microstructure is generalized to take into account the influence of point defect sinks … Abstract A model of intra-grain fission gas bubble growth in U 3 Si 2 coupled with defect microstructure is generalized to take into account the influence of point defect sinks and defect clustering. The dynamics of bubble growth and defect structure properties are studied under different irradiation conditions. The influence of temperature and flux on bubble growth, defect ensemble evolution, and changes in material properties (elastic moduli and thermal degradation factor) are examined in detail. The universality of the bubble size distribution and the crossover of dynamical regimes of bubble growth are studied under various irradiation conditions. It is shown that a change in the dominant (fission gas atom- or vacancy-mediated) mechanism of bubble growth results in a crossover from a parabolic to a sub-parabolic bubble size growth law. The proposed modification of the rate theory model provides more accurate predictions and more detailed insight into fuel performance, especially fission gas behavior in crystalline U 3 Si 2 .

Assessment of the Molten Pool Initial State During Degradation of the Reactor Core with Horizontal Fuel Assemblies Arrangement

2025-06-13

Vladislav Filonov , Dmitriy Fedorov | NTU KhPI Bulletin Power and heat engineering processes and equipment

The paper examines the first approximation in the assessment of the corium initial state in the lower plenum during the in-vessel phase late stage of severe accident of a promising … The paper examines the first approximation in the assessment of the corium initial state in the lower plenum during the in-vessel phase late stage of severe accident of a promising small modular reactor of the ESS-SMART project [1]. The features of the presented analysis are related to both the initial supercritical state of the coolant and the horizontal layout of the core in the form of a system of parallel fuel assemblies with a seven-level coolant inlet system [2]. Existing industry computational system tools are limited in representing the original geometric configuration of the reactor flow path, and in particular the reactor core. Moreover, significant difficulties still arise taking into account the pressure decreasing from a supercritical state to a two-phase state, which seriously limits the performance of estimates using the analytical tool only. In this work, the main principles of sequential equivalent approximation are considered, which allow for a notable simplification of the analysis by isolating stages where only one phenomenon is dominant. For example, the evaluation of decompression parameters and impact loads is the first stage, while the immediate reactor core degradation and the formation of the core molten materials pool is the second one. The third stage is a detailed assessment of the reactor vessel degradation. In this work, attention is focused on the second stage. For this, a fast run model was prepared in the MELCOR 1.8.6 integral code with an equivalent approximation in mass fractions of the "components" in the reactor core flow part. In addition, the conceptual design of the lower plenum of the SCW-SMR reactor is similar to that of typical PWR reactors. The results of the design scenarios made it possible to form a general picture of the reactor core degradation progression at the early phase of a severe accident for the selected initial conditions, as well as to evaluate the characteristics of the accumulated melt in the lower plenum at the late phase of in-vessel severe accident stage. The obtained results are an input set for further evaluation of reactor vessel degradation.

Characterization of Fuel-to-Coolant Heat Transfer During Reactivity-Initiated Accidents Using Tightly Coupled Thermal Hydraulics and Fuel Thermomechanics

2025-06-13

Mason Fox , Jacob P. Gorton , Charles P Folsom +1 more | Nuclear Technology

High-Temperature Structure Design and Integrity Evaluation of Type 316L Stainless Steel Hot Tank

2025-06-13

Yun-Seung Lee , Hyeong-Yeon Lee , Ki-Ean Nam +2 more | Transactions of the Korean Society of Mechanical Engineers A

Determination of elastic-plastic properties of hydride precipitate in uranium matrix by nanoindentation and finite element analysis

2025-06-12

Ruiwen Li , Wen‐Liang Xu , Zhiyuan Wen | Results in Engineering