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 Pressurized water reactor spent nuclear fuel data library produced with the Serpent2 code

Data in Brief ◽  
2020 ◽  
Vol 33 ◽  
pp. 106429
Author(s):  
Zsolt Elter ◽  
Li Pöder Balkeståhl ◽  
Erik Branger ◽  
Sophie Grape
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Data Library

Physico-chemical characterization of a spent UO2 fuel with respect to its stability under final disposal conditions

2014 ◽  
Vol 1665 ◽  
pp. 283-289 ◽  
Author(s):  
Ernesto González-Robles ◽  
Detlef H. Wegen ◽  
Elke Bohnert ◽  
Dimitrios Papaioannou ◽  
Nikolaus Müller ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Chemical Characterization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Final Disposal ◽  
Fuel Rod ◽  
Physico Chemical ◽  
The Stability ◽  
Fission Gas Release

ABSTRACTTwo adjacent fuel rod segments were irradiated in a pressurized water reactor achieving an average burn-up of 50.4 GWd/tHM. A physico-chemical characterisation of the high burn-up fuel rod segments was performed, to determine properties relevant to the stability of the spent nuclear fuel under final disposal conditions. No damage of the cladding was observed by means of visual examination and γ-scanning. The maximal oxide layer thickness was 45 µm. The relative fission gas release was determined to be (8.35 ± 0.66) %. Finally, a rim thickness of 83.7 µm and a rim porosity of about 20% were derived from characterisation of the cladded pellets.

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel—I: Methodology Overview

1999 ◽  
Vol 125 (3) ◽  
pp. 255-270 ◽  
Author(s):  
Dale B. Lancaster ◽  
Emilio Fuentes ◽  
Chi H. Kang ◽  
Meraj Rahimi
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Multiphysics Modeling of Pressurized Water Reactor Fuel Performance

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chungyu ◽  
Yuan Cenxi
Keyword(s):  
Three Dimensional ◽  
Fission Products ◽  
Axial Direction ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Rods ◽  
Reactivity Insertion

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

Validation of lattice physics code STREAM for predicting pressurized water reactor spent nuclear fuel isotopic inventory

2018 ◽  
Vol 120 ◽  
pp. 431-449 ◽  
Author(s):  
Bamidele Ebiwonjumi ◽  
Sooyoung Choi ◽  
Matthieu Lemaire ◽  
Deokjung Lee ◽  
Ho Cheol Shin
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Summary of Results From the Series 2 and Series 3 NNWSI Bare Fuel Dissolution Tests

1987 ◽  
Vol 112 ◽  
Author(s):  
C. N. Wilson
Keyword(s):  
Spent Nuclear Fuel ◽  
Fine Particles ◽  
Fission Products ◽  
Test Site ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Nuclear Waste Storage ◽  
Waste Storage ◽  
Rate Of Increase ◽  
In Series

AbstractThe Nevada Nuclear Waste Storage Investigations (NNWSI) Project is studying dissolution and radionuclide release behavior of spent nuclear fuel in Nevada Test Site groundwater. Specimens prepared from pressurized water reactor (PWR) fuel rod segments were tested for multiple cycles in J-13 well water. The Series 2 tests were run in unsealed silica vessels under ambient hot cell air (25°C) for five cycles for a total of 34 months. The Series 3 tests were run in sealed stainless steel vessels at 25°C and 85°C for three cycles for a total of 15 months. Selected summary results from Series 2 and Series 3 tests with bare fuel specimens are reported.Actinide concentrations tended to saturate and then often decreased during test cycles. Uranium concentrations in later test cycles ranged from 1 to 2 μg/ml in the Series 2 Tests versus about 0.1 to 0.4 μg/ml in Series 3 with the lowest concentrations occurring in the 85°C tests. Formation of a calciumuranium-silicate phase identified as uranophane in the 85°C Series 3 Tests is thought to have limited uranium concentration in these tests. Americium-241, Pu-239 and Pu-240 activities measured in filtered solution decreased to less than 1 pCi/ml in the 85°C tests. Preferential release of fission products Cs, I, Sr and Tc, and activation product C-14, was indicated relative to the actinides. Tc-99 and Cs-137 activities measured in solution after Cycle 1 increased linearly with time, with the rate of increase greater at 85°C than at 25°C. Continuous preferential release of soluble fission products is thought to result primarily from the dissolution of fine particles of fission product phases concentrated on grain boundaries.

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.

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