Physicochemical and structural effects of electrolyte‐induced delamination on polyethylene teraphthalate‐coated steel plates

This research evaluated the in situ physicochemical changes and alterations occurring in an electrolytic chromium coated steel (ECCS), surface protected by polyethylene teraphthalate (PET) copolymer, after inducing a fracture on the coating in an acid acetic‐acetate medium. The delamination was characterized in the front of the failure by means of anodic and cathodic electrochemical mechanisms, and the resistance and degradation of the metal‐polymer composite's substrates were analyzed by means of Raman vibrational spectroscopy. The application of an electrochemical cell to generate in situ delamination, simulating the formation of pores or artificial defects, provided information on the activity inside the substrates of the PET‐coated ECCS composite as a result of the effect of the acetic acid. The anodic delamination mechanism is based on the diffusion of the electrolyte through the metal‐polymer interface and the pre‐existence of pores on the polymer layer. The cathodic delamination mechanism is based on the mechanical action of the gaseous hydrogen as a result of the reduction of H⁺. © 2011 Wiley Periodicals, Inc. J Appl Polym Sci, 2012     

Modern Technology Applied in Advanced BWR (ABWR)

The Advanced Boiling Water Reactor (ABWR) has been developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced radiation exposure and radwaste. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. The ABWR incorporates the best proven features from BWR designs in Japan, the United States and Europe. The many new features are seen to provide superiority in terms of performance characteristics and economics relative to current LWR designs. The Tokyo Electric Power Co., Inc. recently announced the selection of General Electric Co., Hitachi, Ltd. and Toshiba Corp. to design and construct two lead Advanced Boiling Water Reactors as Unit 6 and 7 at the Kashiwazaki Kariwa Nuclear Power Station. Construction is scheduled for the early 1990's, and commercial operation planned for 1996 for Unit 6 and 1998 for Unit 7.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

A feasibility assessment of the use of nanofluids to enhance the in-vessel retention capability in light-water reactors

Nanofluids, colloidal dispersions of nanoparticles, exhibit a substantially higher critical heat flux (CHF) compared to water. As such, they could be used to enhance the in-vessel retention (IVR) capability in the severe accident management strategy implemented by certain light-water reactors. It is envisioned that, at normal operating conditions, the nanofluid would be stored in dedicated storage tanks, which, upon actuation, would discharge into the reactor cavity through injection lines. The design of the injection system was explored with risk-informed analyses and computational fluid dynamics. It was determined that the system has a reasonably low failure probability, and that, once injected, the nanofluid would be delivered effectively to the reactor vessel surface within seconds. It was also shown analytically that the increase in decay power removal through the vessel using a nanofluid is about 40%, which could be exploited to provide a higher IVR safety margin or, for a given margin, to enable IVR at higher core power. Finally, the colloidal stability of a candidate alumina-based nanofluid in an IVR environment was experimentally investigated, and it was found that this nanofluid would be stable against dilution, exposure to gamma radiation, and mixing with boric acid and lithium hydroxide, but not tri-sodium phosphate.     

Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina Nuclear Power Plant

Stress corrosion cracks have been discovered in Group Distribution Headers (GDH) at the Ignalina and Chernobyl Nuclear Power Plants. This increases the probability that a guillotine pipe break can occur that creates a whipping pipe (GDH) with the potential to damage surrounding structures—i.e. adjacent GDH and its attached piping or adjacent reinforced concrete compartment wall. The GDH is the most important component for reactor safety in case of an accident. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the ECSS into the GDH.Presented in this paper is the transient analysis of a Group Distribution Header following a guillotine break at the blind end of the header. Using a very conservative force loading function, the transient response of a whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is obtained using finite element methodology.The results of the study, assuming that the impacted GDH does not suffer stress corrosion cracking, indicate that the structural integrity of the compartment should be maintained and failure should not propagate from GDH to GDH.     

Application of Unavailability Analysis Method to Determining Periodic Test Procedure of Emergency Core Cooling System

The test interval and the test procedure are the main factors that are related to the reliability of the emergency core cooling system in a nuclear power plant. A method is proposed to specify the allowable ranges of the test intervals for the minimal cut sets in the emergency core cooling system; lower and upper limits of test intervals are selected to minimize unavailability and to assure the unavailability goal respectively. A method is also proposed to determine test procedure. All patterns of test procedure for the emergency core cooling system are generated in the allowable ranges of test intervals, and the test procedure is selected to maximize the index {minus log. of (unavailability over unavailability goal), all devided by man-hours} for the purpose of both reliability increase and man-hours decrease.     

Experimental Study on Difference in Reflood Core Heat Transfer among CCTF,FLECHT-SET and Predicted with FLECHT Correlation

It has been reported that the core heat transfer coefficients measured in the CCTF tests, which were conducted under the conditions expected to appear during the refiooding period in a PWR, can not be predicted well with the FLECHT correlation, which has been used in the safety evaluation. In order to investigate the reason for this, a CCTF test was conducted under the typical FLECHT-SET experimental conditions. Investigating results from both tests, the following has been clarified:

The FLECHT correlation can not describe the heat transfer for the refiooding situations with the initial Accumulator injection period, which is expected to appear in a PWR, and gives much lower values than the measured. The core heat transfer in the FLECHT-SET is similar to that in the CCTF, and they are well predicted with the Murao-Sugimoto correlation. When there is some core radial power distribution, which strongly affects the heat transfer in a large vScale core, the heat transfer coefficients in the CCTF can be well predicted with taking account of this effect in addition to the Murao-Sugimoto correlation.     

Integral analysis of large LOCA beyond design basis accident scenario in RBMK-1500

The integral analysis of severe accident scenario for RBMK-1500 was performed using combined approach with RELAP5, RELAP/SCDAPSIM, ASTEC and COCOSYS codes. The performed analysis covered response of the reactor core, the reactor cooling system and the confinement. There were performed several analyses: the first analysis assumed that operators take no action or their actions are not successful to provide the coolant injection to the reactor core; the other analyses were performed to investigate the accident management measures to restore the core cooling at different temperatures of the reactor core. The results of performed analyses showed that the operators have ∼5 h before the ruptures of fuel claddings occur and ∼8 h before the onset of exothermic steam-zirconium reaction. The coolant injection to the reactor core should be restored as soon as possible in order to prevent high hydrogen concentrations in the confinement and significant release of the fission products to the environment.     

Effect of Gas Atoms on Recovery of Work Hardening in Al-Li Alloys

In the first report of this study, dealing with CCFL and CCFL breakdown phenomena associated with the injection of emergency core cooling spray water into upper plenum during refill-reflood phase of a BWR LOCA, the following tests results were obtained.

The injected water maintained two-phase pool across the top of entire core after CCFL breakdown. The pool level oscillated near spray elevation. The objective of this paper is to clarify the mechanism of these phenomena, evaluating steam and spray flow effects on CCFL breakdown.

It is found that when spray flow rate was slightly larger than the CCFL drainage deter- mined by core steam flow, pool maintained at some constant level near spray elevation, after CCFL breakdown. On the other hand, when spray flow was appreciably larger than CCFL drainage, pool level slowly oscillated. The oscillation was caused by significant changes in steam condensation rate, and the corresponding subcooling penetration into the fuel bundles, when the pool level passed the spray elevation. The TRAC-BD1 analysis of test results suggested the small sector wall effect of test apparatus on CCFL breakdown phenomena.     

An Analytical Model for Displacement Velocity of Liquid Film on a Hot Vertical Surface

The downward progress of the advancing front of a liquid film streaming down a heated vertical surface, as it would occur in emergency core cooling, is much slower than in the case of ordinary streaming down along a heated surface already wetted with the liquid. A two-dimensional heat conduction model is developed for evaluating this velocity of the liquid front, which takes account of the heat removal by ordinary flow boiling mechanism.

In the analysis, the maximum heat flux and the calefaction temperature are taken up as parameters in addition to the initial dry heated wall temperature, the flow rate and the velocity of downward progress of the liquid front. The temperature profile is calculated for various combinations of these parameters. Two criteria are proposed for choosing the most suitable combination of the parameters. One is to reject solutions that represent an oscillating wall temperature distribution, and the second criterion requires that the length of the zone of violent boiling immediately following the liquid front should not be longer than about 1 mm, this value being determined from comparisons made between experiment and calculation.

Application of the above two criteria resulted in reasonable values obtained for the calefaction temperature and the maximum heat flux, and the velocity of the liquid front derived therefrom showed good agreement with experiment.