scholarly journals Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage

2021 ◽  
Vol 1 ◽  
pp. 7-8
Author(s):  
Mara Marchetti ◽  
Michel Herm ◽  
Tobias König ◽  
Simone Manenti ◽  
Volker Metz
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Reactor Core ◽  
Fission Products ◽  
Alpha Decay ◽  
Irradiation Damage ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
Dry Storage ◽  
Inner Surface

Abstract. After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

Neutron Physics Characterization & Optimization Analysis of the ACPR100 Small Modular Reactor

2021 ◽  
Author(s):  
Songyang Liu ◽  
Xiang Wang
Keyword(s):  
Life Cycle ◽  
Reactor Core ◽  
Fission Products ◽  
Negative Temperature ◽  
Design Parameters ◽  
Monte Carlo Code ◽  
Time Step ◽  
Optimization Analysis ◽  
Insertion Depth ◽  
Pressurized Water

Abstract The ACPR100 is a small modular pressurized water reactor design proposed by China General Nuclear Power Corporation which integrates most important components of the reactor into one pressure vessel. This paper aims to model the reactor core by Monte Carlo code Serpent. Firstly, the steady-state characterization and optimization analysis within different temperature, pin-pitch and other design parameters are performed. Secondly, the loading pattern of fuel assemblies with Gd-doped fuel rods was assessed. Thirdly, based on the reference research, the temperature coefficients of fuel and coolant are calculated. The influence of control rod insertion depth was simulated additionally. Besides, to show the life cycle and the change of inventory directly, we performed burnup calculation based on pre-defined time step and discussed the fission products including radioactive minor actinides and radiopharmaceutical isotopes during the life cycle. The results show that the ACPR100 achieved more stable performance using high abundance boric acid, and the negative temperature reactivity coefficient is sufficient to maintain the stability of the reactor operation, but the ACPR100 is not suitable for massive production of radiopharmaceutical isotopes.

Study of the release of the fission gases (Xe and Kr) and the fission products (Cs and I) under anoxic conditions in bicarbonate water

2015 ◽  
Vol 1744 ◽  
pp. 35-41 ◽  
Author(s):  
Ernesto González-Robles ◽  
Markus Fuß ◽  
Elke Bohnert ◽  
Nikolaus Müller ◽  
Michel Herm ◽  
...  
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Collaborative Project ◽  
Fuel Rod ◽  
Release Ratio

ABSTRACTFor safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.

scholarly journals Nuclear Fuel Cycle with Minimized Waste

2019 ◽  
pp. 30-35
Author(s):  
V. Moiseenko ◽  
S. Chernitskiy
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
Nuclear Fuel Cycle ◽  
Fission Neutron ◽  
Fuel Composition ◽  
Monte Carlo Code ◽  
Metallic Fuel

A uranium-based nuclear fuel and fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a ‘balanced’ fuel only uranium-238 content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The major fissionable component of the fuel is plutonium is chosen. This makes it possible to abandon the use of uranium-235, whose reserves are quickly exhausted. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. This feature simplifies maintaining the closed nuclear fuel cycle and provides its periodicity. In the fuel balance calculations, nine isotopes of uranium, neptunium, plutonium and americium are used. This number of elements is not complete, but is quite sufficient for calculations which are used for conceptual analysis. For more detailed consideration, this set may be substantially expanded. The variation of the fuel composition depending on the reactor size is not too big. The model accounts for fission, neutron capture and decays. Using MCNPX numerical Monte-Carlo code, the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components. In this way, it can be assumed that transuranic elements will constantly return to such a reactor, and only fission products will be sent to storage. This will significantly reduce the radioactivity of spent nuclear fuel. It is important that the storage time for the fission products is much less than for the spent nuclear fuel, just about 300 years.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

In-Service Inspection of Extended Dry Storage of Spent Nuclear Fuel, Part I: Crawler Technology Development

2021 ◽  
Author(s):  
Ryan M. Meyer ◽  
Jeremy Renshaw ◽  
Jamie Beard ◽  
Jon Tatman ◽  
Matt Keene ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Technology Development ◽  
Storage System ◽  
Dry Storage ◽  
Majority Population ◽  
The Usa ◽  
Extended Operation ◽  
Interim Storage ◽  
Physical Access

Abstract This paper describes development and demonstration of remote crawling systems to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA. Specifically, this work relates to robotic crawler developments for “canister” based DSS systems, which form the majority population of DSSs in the USA for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond their initial licensed or certified terms. Challenges with performing the periodic examinations are associated with physical access to the canister surface, which is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics. Motivations for performing periodic examinations and developing robotic crawlers for performing those examinations remotely will be presented, and several activities to demonstrate robotic crawlers for different DSS systems are summarized.

Inservice Inspection of Extended Dry Storage of Spent Nuclear Fuel, Part II: NDE/Sensor Technology Development and Codification

2021 ◽  
Author(s):  
Ryan M. Meyer ◽  
Jeremy Renshaw ◽  
Kenn Hunter ◽  
Mike Orihuela ◽  
Jim Stadler ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Technology Development ◽  
Storage System ◽  
Sensor Technology ◽  
Nondestructive Examination ◽  
Dry Storage ◽  
Safety Function ◽  
Majority Population ◽  
The Usa

Abstract This paper describes development and demonstration of nondestructive examination (NDE) technologies to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA to verify continued safe operation and that the canister confinement is intact and performing its intended safety function. Specifically, this work relates to NDE technology development for “canister” based DSS systems, which form the majority population of DSSs in the USA for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond the initial license or certified term. Physical access to the canister surface is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics related to inspections. Several activities to demonstrate NDE technologies for the inspections of different DSS systems are summarized.

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