Experimental study of the Dissolution of spent fuel at 85°C in Natural Ground Water

1986 ◽  
Vol 84 ◽  
Author(s):  
Charles N. Wilson ◽  
Henry F. Shaw
Keyword(s):  
Environmental Protection Agency ◽  
Fused Silica ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Waste Package ◽  
Fuel Rod ◽  
Dissolution Tests ◽  
J 13 ◽  
End Caps

AbstractSemi-static dissolution tests using pressurized water reactor spent fuel rod segments and NNWSI reference J-13 well water in sealed stainless steel vessels at 85°C are being conducted in support of the Waste Package Task of the NNWSI Project. Test specimens include: bare fuel plus the empty cladding hulls, fuel rod segments with artificially induced cladding defects and water-tight end caps, and undefected fuel rod segments with water-tight end caps. The test conditions approximate those expected in the proposed NNWSI Project repository when the waste package has cooled sufficiently to allow water to enter a breached container and contact the fuel rods, some of which may exhibit various degrees of cladding failure. Periodic solution samples (unfiltered and filtered) were analyzed for most radionuclides for which cumulative release limits are listed by the U.S. Environmental Protection Agency. Results from the first six-month cycle of the 85° C tests are presented and are compared with results from the first cycle of a previous test series run at 25° in fused silica test vessels.

scholarly journals Derivation of a Waste Package Source Term for NNWSI from the Results of Laboratory Experiments

1985 ◽  
Vol 50 ◽  
Author(s):  
Virginia M. Oversby ◽  
Charles N. Wilson
Keyword(s):  
Laboratory Experiments ◽  
Source Term ◽  
Yucca Mountain ◽  
Maximum Amount ◽  
Well Water ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
J 13 ◽  
Fractional Release

AbstractResults are presented for the dissolution of Turkey Point pressurized water reactor (PWR) spent fuel in J-13 well water at ambient hot cell temperatures. These results are compared with those previously obtained on Turkey Point fuel in deionized water, on H. B. Robinson PWR fuel in J-13 water, and by other workers using various fuels in dilute bicarbonate groundwaters. A model is presented that represents the conditions under which maximum dissolution of spent fuel could occur in a repository sited at Yucca Mountain, Nevada. Using an experimentally determined upper limit of 5 mg/l for uranium solubility in J-13 water, a fractional release rate of 6.4 × 10−8 per year is obtained by assuming that all water entering the repository carries away the maximum amount of uranium.

scholarly journals Results from Long-Term Dissolution Tests Using Oxidized Spent Fuel

1990 ◽  
Vol 212 ◽  
Author(s):  
Charles N. Wilson
Keyword(s):  
Test Site ◽  
Fused Silica ◽  
Spent Fuel ◽  
Test Cycle ◽  
Release Rates ◽  
Fuel Oxidation ◽  
Dissolution Tests ◽  
J 13 ◽  
Fractional Release

ABSTRACTTwo semi-static dissolution tests using oxidized PWR spent fuel specimens are being conducted under ambient hot cell conditions in Nevada Test Site J-13 well water and unsealed fused silica vessels. The test specimens were oxidized at 250°C in air to bulk oxygen-to-metal (O/M) values of 2.21 and 2.33. Following an initial 191-day test cycle, the specimens were restarted in fresh J-13 water for a second long-term test cycle. Results through the first 40 months of Cycle 2 are compared with results from similar tests at 25°C and 85°C using unoxidized spent fuel specimens.Increased concentrations of U, Am, Cm and Np were measured in 0.4- μm filtered samples from the oxidized fuel tests compared to the unoxidized fuel tested at 25°C; Pu concentrations were not affected by the fuel oxidation state. Most of the Am and Cm, and a portion of the Pu, measured in 0.4-μm filtered samples was removed by 2-nm filtration. Fission product release results were normalized to specimen inventories and reported as fractional release. No attempt was made to normalize the data to surface area. Initial 99Tc release was greatly increased, and prolonged increases in the fractional release rates of 99Tc and 129I occurred as a result of fuel oxidation. Fractional release rates for 137Cs and 90Sr from oxidized fuel eventually decreased to levels similar to those observed with unoxidized fuel after equivalent testing times, suggesting that matrix dissolution rates normalized to fuel mass were not increased as a result of oxidation.

Development of a Thermal Analysis Model for a Nuclear Spent Fuel Storage Cask and Experimental Verification With Prototype Testing

1989 ◽  
Vol 111 (4) ◽  
pp. 647-651 ◽  
Author(s):  
J. Y. Hwang ◽  
L. E. Efferding
Keyword(s):  
Thermal Analysis ◽  
Storage Period ◽  
Department Of Energy ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Analysis Model ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage

A thermal analysis evaluation is presented of a nuclear spent fuel dry storage cask designed by the Westinghouse Nuclear Components Division. The cask is designed to provide passive cooling of 24 Pressurized Water Reactor (PWR) spent fuel assemblies for a storage period of at least 20 years at a nuclear utility site (Independent Spent Fuel Storage Installation). A comparison is presented between analytical predictions and experimental results for a demonstration cask built by Westinghouse and tested under a joint program with the Department of Energy and Virginia Power Company. Demonstration testing with nuclear spent fuel assemblies was performed on a cask configuration designed to store 24 intact spent fuel assemblies or canisters containing fuel consolidated from 48 assemblies.

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

The Effect of MA Transmutation in the PWR on Fuel Cycle

2017 ◽  
Author(s):  
Haoyang Yu ◽  
Bin Liu ◽  
Wenxin Zhang ◽  
Jin Cai
Keyword(s):  
Fuel Cycle ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Mcnp Program ◽  
And Control ◽  
Control Rods ◽  
Effective Multiplication

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

Robust design of fuel rods against potential structural damage by flow-induced vibration

2019 ◽  
Vol 33 (01n03) ◽  
pp. 1940008
Author(s):  
Huan-Huan Qi ◽  
Nai-Bin Jiang ◽  
Yi-Xiong Zhang ◽  
Zhi-Peng Feng ◽  
Xuan Huang
Keyword(s):  
Structural Damage ◽  
Flow Distribution ◽  
Fretting Wear ◽  
Leaf Spring ◽  
Wear Depth ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Vibration Displacement ◽  
Pressurized Water ◽  
Fuel Rod

We studied the flow-induced vibration (FIV) and fretting wear of fuel rod with grid relaxation. According to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. By combining the correlation of PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The grids may relax due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of grid clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the responses amplitude of turbulent excitation at the bottom and top of the fuel rod are larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure on the rod wear depth depicts basically the same trend as on the maximum FIV amplitude.

Tripoli-4 Gamma-Ray Dose Calculation for Spent PWR Fuels

2013 ◽  
Author(s):  
Yi-Kang Lee ◽  
Kabir Sharma
Keyword(s):  
Gamma Ray ◽  
Dose Calculation ◽  
Radiation Shielding ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Heterogeneous Model ◽  
Transport Code ◽  
Fuel Pin ◽  
Monte Carlo Transport

The gamma-ray dose calculation is essential for the radiation shielding of pressurized water reactor (PWR) spent fuels. Homogenization modeling of fuel pin lattices for typical PWR spent fuel pins is regularly applied on the radiation protection calculation of gamma-ray dose in an air medium. However, depending on the size of the homogenized lattice and the location of the detectors, under-estimation or over-estimation of the gamma-ray dose due to the homogenization modeling can be obtained with respect to the detailed heterogeneous model. In previous published results from MCNP-4A and 4C calculations on gamma-ray dose from spent PWR fuel pins, very different homogeneous to heterogeneous (Hom/Het) ratios were reported. In this study these Hom/Het ratios have been re-evaluated and benchmarked by using the TRIPOLI-4 Monte Carlo transport code. The new TRIPOLI-4 mesh tally capabilities have also been applied to calculate the radial and axial gamma-ray dose distribution. With the recently upgraded TRIPOLI-4 display tool, the dose rate maps and the isodose rate curves around a spent PWR fuel assembly have been established.

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.

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