Toshiari Saegusa

The Long-Term Performance of Concrete in Nuclear Applications

The Idaho National Laboratory investigated long-term concrete performance for nuclear applications. Scientists searched the open literature for information on the effects of heat and radiation. They also examined the concrete shield of the Pacific Sierra Nuclear Ventilated Storage Cask (VSC-17) for concrete deterioration and loss of shielding. The literature search revealed that low doses, <109 n/cm2 or <1011 Rad gamma, of radiation over periods less than 50 years do not appear to significantly affect concrete. Exposure over 100 years was not studied. The effects of higher doses of radiation are not as clear. Generally, the threshold of degradation is 95°C, and degradation increases with increasing temperature and time. In 15 years of VSC-17 use, there is no apparent effect from environment, radiation, or temperature that has adversely affected the shielding or structural functions, although there has been some minor cracking of the concrete.© 2005 ASME

Experimental Studies on Safety of Dry Cask Storage Technology of Spent Fuel Allowable Temperature of Cladding and Integrity of Cask under Accidents

The purpose of this report is to publish results on a comprehensive review of research subjects on safety of the dry cask storage and to publish a part of the research results. (1) A method to calculate the maximum allowable temperature of fuel cladding was proposed, based on experiments using irradiated and non-irradiated Zircaloy tube specimens, in order to prevent excessive creep deformation of fuel cladding in dry cask. (2) Under an accidental test of “Cask drop onto concrete floor”, the cask integrity was demonstrated using full-scale casks (100-t class). (3) Under an accidental test of “Building collapse and drop of heavy objects onto cask”, the cask integrity was demonstrated using a full-scale cask (100-t class). (4) As to possibility of cask toppling by earthquake, analytical and experimental results using a scale model cask showed that the cask will not be toppled by earthquake.

Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads Due to Aircraft Engine Crash

In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters and seismic tests subjected to strong earthquake motions. Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001. This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are developed and calculated. Main criteria for estimating the maximum leakage rate for the lid metallic seal system are no loss of the pre-stress of the lid bolts, no appearance of the plastic region between the metal seal flanges, and no large relative deformation of the lid seals. Finally, in both cases, the low leakage rate for the metal cask lid closure system under the impulsive loads due to aircraft engine crash will be proved thoroughly.Copyright

Transport and storage of spent nuclear fuel

This chapter introduces the amount of spent fuel generation and storage in the world and the characteristics of spent fuel. An overview of the historical development of spent fuel storage technologies is provided and some examples of the storage casks are introduced. Current topics on transport and storage of spent fuel are described, including issues of long-term storage, long-term containment of metal gaskets for metal casks, interactions between transport and storage on containment, stress corrosion cracking of canisters, and a holistic approach to assure transport and storage safety of metal casks.

Japanese Codes and Standards for Dual Purpose Metal Casks for Spent Nuclear Fuel

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied.The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME.JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added.The Code has differences from that for NPP components with considerations to DPC characteristics;- The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1).- Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function.- Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages.- Stress limits for earthquakes during storage are specified.- Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels.DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.Copyright

Development of Salt Particle Collection Device to Prevent SCC of Canisters

A natural cooling system is economical for removing the decay heat from casks at interim storage facilities of spent nuclear fuel. At storage facilities of concrete casks built near the seashore, the air including the sea salt particles comes into the concrete casks and may cause Stress Corrosion Cracking (SCC) to the canister made of welded stainless steel plates. In order to prevent SCC on the canister, it is necessary to keep the density of salt on the surface of the canister smaller than the threshold which causes SCC. In this study, the authors propose a salt particle collection device with a low flow resistance which doesn’t block the air inlet of the storage building. The salt particle collection device is installed at the inlet and composed of a duct with multiple trays where the water is filled. When the air including sea salt particles comes through the duct, it collides with the surface of the water in the trays, and the part of sea salt particles in the air dissolves in the water. Therefore, the salt carried into the building by the air is reduced. The salt water in the trays is discharged out of the building by overflowing. The device has the following characteristics. a) Because of the low flow resistance, the device doesn’t block the inflow of the air which needs for removing the heat from casks. b) Rainwater may be usable for the water used in the device. c) Because of the simple structure, the maintenance is easy. The authors conducted the experiments using the device in the laboratory and outdoors near the seashore. The obtained results are as follows: (1)The pressure loss of the device is 1/7 of that of a filter used in a forced cooling system and the efficiency of salt particle collection is approximately 24%. (2)The efficiency of salt particle collection in the outdoor tests is larger than that in the laboratory tests. By further experimental study, the authors will develop the device with lower pressure loss and higher efficiency of salt particle collection.Copyright

Recent Development of Code Case on Use of Ductile Cast Iron for Transport and Storage Cask for Spent Nuclear Fuel

Code Case N-670 “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-1992 for Transport Containments, Section III, Division 3” which permits use of ductile cast iron for transport containments of spent nuclear fuel was revised to the Code Case N-670-1, “Use of Ductile Cast Iron Conforming to ASTM A874/A 874M-98 or JIS G5504-2005 for Transport and Storage Containments, Section III, Division 3”. Items revised were as follows: (a) Scope was expanded to use for transport and storage, and changed to conform year edition of JIS G5504, (b) The elongation requirement was deleted form the code case to reflect the change of year edition of JIS G5504, (c) Temperature condition of −40 °C was clearly provided for fracture toughness test, (d) Design fatigue curve was re-established, (e) External pressure chart was re-established. Technical basis of the revised code case are described in this paper.Copyright

Atmospheric Corrosion Control on the basis of Radiation Induced Surface Activation

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Fracture Toughness of Ductile Cast Iron and Applicability of Fracture Mechanics to DCI Casks

The JSME Cask Code, Rules for Construction of Metallic Casks, has a scope of using three types of casks fabricated by stainless steels, forged low alloy steels and ductile cast irons (DCI). On the other hand, the use of DCI for Cask material is not within the scope of the ASME Code, Section III, Division 3. In this paper, the fracture toughness of DCI for JSME cask code was compared with those of quenched and tempered low alloy steels. Furthermore, applicability of fracture mechanics to DCI casks is demonstrated by the results of fracture tests of reduced scale model casks. The results deny criticism arising from the fact that the DCI is not homogeneous material due to precipitation of spheroidal graphite particles for the application of the fracture mechanics.Copyright