scholarly journals THERMAL ANALYSIS OF DRY STORAGE AND TRANSPORTATION CASKS CASTOR USING COBRA-SFS

2020 ◽  
Vol 28 ◽  
pp. 32-41
Author(s):  
Martin Ševeček ◽  
Mojmír Valach ◽  
Chih-Hao Lee
Keyword(s):  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Construction Materials ◽  
Safety Evaluation ◽  
Fission Products ◽  
The Czech Republic ◽  
Dry Storage ◽  
State Office ◽  
The World ◽  
Residual Heat

State Office for Nuclear Safety is a regulator body in the Czech Republic handling also licensing of storage and transportation casks (SCs) of the spent nuclear fuel (SNF). One of the main concerns for the dry storage of SNF is the safe removal of the residual heat resulting from the decay of fission products, actinides, and activated construction materials. There are many storage systems in use around the world with different storage configuration, fuel designs and boundary conditions that need to be evaluated. Two thermal models of SNF storage casks CASTOR 440/84 and CASTOR 440/84M were developed using the COBRA-SFS code, which is an internationally accepted and validated code used for licensing of dry storage casks and safety evaluation. The casks are in operation at the Nuclear Power Plant Dukovany for dry storage of the WWER-440 SNF. Both models were validated and used for evaluation of several problems in different configuration and particularly for licensing purposes at the SUJB.

Electrochemical decontamination of irradiated nuclear graphite from corrosion and fission products using molten salt

2021 ◽  
Author(s):  
Tatiana Grebennikova ◽  
Abbie N Jones ◽  
Clint Alan Sharrad
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Waste Management ◽  
Molten Salt ◽  
Nuclear Power ◽  
Fission Products ◽  
Nuclear Graphite ◽  
Irradiated Graphite ◽  
The World ◽  
The Uk

Irradiated graphite waste management is one of the major challenges of nuclear power-plant decommissioning throughout the world and significantly in the UK, France and Russia where over 85 reactors employed...

scholarly journals On the prospects of using metal hydrides in nuclear energy

2020 ◽  
Vol 21 (4) ◽  
pp. 308-311
Author(s):  
M.I. Vlasenko ◽  
◽  
M.M Korotenko ◽  
S.L. Lytvynenko ◽  
D.V. Pyshna ◽  
...  
Keyword(s):  
Power Plant ◽  
Neutron Irradiation ◽  
Nuclear Power ◽  
Nuclear Energy ◽  
Spent Nuclear Fuel ◽  
Experimental Studies ◽  
Metal Hydrides ◽  
Protective Properties ◽  
Dry Storage ◽  
The Usa

The results of analytical and experimental studies of neutron-protective properties of a number of domestic materials and of the SWX-277 material (manufactured in the USA) are reported. SWX-277 is employed for protection against neutron irradiation in containers for dry storage of spent nuclear fuel in the Zaporizhzhya Nuclear Power Plant (NPP). The results of studies have confirmed the high protective properties of some domestic materials (in particular, titanium hydride), and the perspective of their wider use for protection against neutron irradiation.

Distribution of Palladium During Spent Fuel Reprocessing

2003 ◽  
Author(s):  
Yu. Pokhitonov ◽  
V. Romanovski ◽  
P. Rance
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Platinum Group Metals ◽  
Spent Fuel ◽  
Platinum Metals ◽  
Spent Fuel Reprocessing ◽  
Platinum Group ◽  
Fuel Reprocessing

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile and fertile isotopes and facilitate disposal of waste elements. Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants (NPPs,) there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. Given current predictions for nuclear power generation, it is predicted that the quantities of palladium to be accumulated by the middle of this century will be comparable with those of the natural sources, and the quantities of rhodium in spent nuclear fuel may even exceed those in natural sources. These facts allow one to consider spent nuclear fuel generated by NPPs as a potential source for creation of a strategic stock of platinum group metals. Despite of a rather strong prediction of growth of palladium consumption, demand for “reactor” palladium in industry should not be expected because it contains a long-lived radioactive isotope 107Pd (half-life 6,5·105 years) and will thus be radioactive for a very considerable period, which, naturally, restricts its possible applications. It is presently difficult to predict all the areas for potential use of “reactor” palladium in the future, but one can envisage that the use of palladium in radwaste reprocessing technology (e.g. immobilization of iodine-129 and trans-plutonium elements) and in the hydrogen energy cycle may play a decisive role in developing the demand for this metal. Realization of platinum metals recovery operation before HLW vitrification will also have one further benefit, namely to simplify the vitrification process, because platinum group metals may in certain circumstances have adverse effects on the vitrification process. The paper will report data on platinum metals (PGM) distribution in spent fuel reprocessing products and the different alternatives of palladium separation flowsheets from HLW are presented. It is shown, that spent fuel dissolution conditions can affect the palladium distribution between solution and insoluble precipitates. The most important factors, which determine the composition and the yield of residues resulting from fuel dissolution, are the temperature and acid concentration. Apparently, a careful selection of fuel dissolution process parameters would make it possible to direct the main part of palladium to the 1st cycle raffinate together with the other fission products. In the authors’ opinion, the development of an efficient technology for palladium recovery requires the conception of a suitable flow-sheet and the choice of optimal regimes of “reactor” palladium recovery concurrently with the resolution of the problem of HLW partitioning when using the same facilities.

Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel

2006 ◽  
Author(s):  
Masumi Wataru ◽  
Hisashi Kato ◽  
Satoshi Kudo ◽  
Naoko Oshima ◽  
Koji Wada ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Sea Salt ◽  
Storage Facility ◽  
Japan Society ◽  
Spent Fuel ◽  
Dry Storage ◽  
Storage Methods ◽  
Interim Storage

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility (Figure 1). Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack.

Neutronic Analysis for VVER-440 Type Reactor Using PARCS Code

2018 ◽  
Author(s):  
Luca Ratti ◽  
Guido Mazzini ◽  
Marek Ruščák ◽  
Valerio Giusti
Keyword(s):  
Czech Republic ◽  
Cross Sections ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Models ◽  
Reactor Core ◽  
Sensitivity Study ◽  
The Czech Republic ◽  
State Office ◽  
Scale Models

The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios. This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE models of the VVER-440 fuel assemblies have been created in order to produce collapsed and homogenized cross sections necessary for the study with PARCS of the whole VVER-440 reactor core. The sensitivity study of the suitable energy threshold to be adopted for the preparation with SCALE of collapsed two energy-group homogenized cross sections is also discussed. Finally, the results obtained with PARCS core model are compared with those reported in the VVER-440 Final Safety Report.

Research on Estimating Methods and Application of Fuel Rods Defect

2017 ◽  
Author(s):  
Huaibin Li ◽  
Yaru Fu ◽  
Lanfang Mao ◽  
Qiliang Mei
Keyword(s):  
Radiation Dose ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Fission Products ◽  
Fuel Management ◽  
Power Station ◽  
Fuel Rods ◽  
Reactor Coolant ◽  
The World

When fuel rods have defects, the fission products in the fuel rods will come out and enter the reactor coolant through the defects. The release of the fission products will result in the increase of radiation dose, the indeterminacy of the fuel management and will influence the economics and safety of the nuclear power station. Based on the analysis of the typical nuclides activities in reactor coolant, the evaluation of the defect fuel rods can be realized. This paper studied the related analysis around the world and determined the methods to evaluate the number, the defect type (open or tight) and the burn-up of the defect fuel rods. The evaluation method of this paper can be used to evaluate the defects of fuel rods, and can provide valuable information for the fuel management and dose analysis, and also can be a useful technical support to the operation of nuclear power station.

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.

scholarly journals SPENT NUCLEAR FUEL DRY STORAGE: STATUS AND CURRENT ISSUES

2019 ◽  
pp. 80-82
Author(s):  
M.V. Yatsenko ◽  
A.V. Korolyov
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Operating Experience ◽  
Dry Storage ◽  
Long Term Storage ◽  
Open Questions ◽  
The World ◽  
The Status ◽  
Term Storage

Among all the options of spent nuclear storage, dry storage seems to be the most expedient method for the nearest future all over the world. Despite the economic attractiveness and the presence of a positive operating experience, there are still open questions. In this article author investigates the status of dry storage, gives a review of world experience and current issues that need to be solved to ensure safe long-term storage of spent nuclear fuel.

Using CFD to Improve the Thermal Characteristic Analysis Methodology of Dry Storage System for Chinshan Nuclear Power Plant

2010 ◽  
Author(s):  
Yung-Shin Tseng ◽  
Jong-Rong Wang ◽  
Chi-Hung Lin ◽  
Chunkuan Shin ◽  
F. Peter Tsai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Thermal Characteristic ◽  
Life Extension ◽  
Fuel Management ◽  
Dry Storage

Chinshan Nuclear Power Plant (CSNPP) is a two-unit BWR4 plant with 1804MWt power per unit. Taipower Co., the owner of the plant is preparing the life extension procedure to extend the CSNPP operation time. In order to meet the life extension requirement, many issues need to be solved before life extension licensing, such as the spent nuclear fuel management, structure aging etc. For the spent nuclear fuel management, ROC Atomic Energy Council (ROCAEC) certified method is employed to analyze the thermal behaviors of Dry Storage System (DSS). This method uses ANSYS coupled with RELAP5-3D to solve the thermal characteristic and successfully accomplish the licensing procedure of the Chinshan Nuclear Dry Storage Project. However, further validation results demonstrate that the coupled method still exists uncertainty and deficiency. In this study, a new Computational Fluid Dynamics (CFD) numerical model for spend nuclear fuel (NSF) dry storage system (DSS) has been developed to improve the accuracy of DSS thermal analysis results. Its accuracy has been validated by comparing the temperature predictions with the experimental results of VSC-17 DSS. It has been found that the thermal behaviors and physical phenomena in the DSS could be predicted with good agreement for the measurements. Moreover, the uncertainty and reasonableness of results in previous method can be improved by the new thermal analyses methodology.

Sign in / Sign up

Export Citation Format

Share Document