Membrane Covered Probes for O2 and H2 Determination - Diffusion through Liquid Films and Solid Membrane in Practice

In the era of the expansion of hydrogen use, its concentration measurement becomes more important. We further focus on one of the H2 concentration measurement purposes, where the hydrogen diffusion in a solid membrane and in a liquid electrolyte play the key role. To keep optimal process conditions in the primary cooling circuit of nuclear power plants, various chemical species are dosed in. Among the species the concentration of which is monitored in primary coolant, belong oxygen and hydrogen. While plenty of companies offer oxygen sensors suitable for the measurement in the primary coolant, the hydrogen sensor, really selective to H2 concentration, is offered by only one company. It is worth, therefore, accomplishing the development of a hydrogen sensor, which began at UCT Prague in the 1990's and, after several successful measurements in nuclear power plant, interrupted due to fateful events in the research team. We introduce here the results of the first part of contemporary work of the Mass Transfer Laboratory based on new technologies but using the experience from 1990's. Having at disposal modern functional samples to measure both oxygen and hydrogen concentrations, we verified a fair long-term stability of the sensors and, further, we would like to cooperate with an industrial partner to finalize the development of prototypes and start the production of monitoring units.

Simulation of VVER-1000 Guillotine Large Break Loss of Coolant Accident Using RELAP5/SCDAPSIM/MOD3.5

The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

ANALYSIS OF RELIABILITY OF AUTOMATIC CORE PROTECTION FUNCTION OF THE REACTOR V-412 IN RESPONSE TO LOCAL PARAMETERS: MAXIMUM LINEAR POWER, DEPARTURE FROM NUCLEATE BOILING RATIO

The research subject is finding an engineering solution for V-412 core automatic protection during operation in both steady-state and transient conditions within ICIS using local parameters (i.e. maximum linear power, departure from nucleate boiling ratio). Such engineering solution will be implemented by safety system software-hardware (PTK-Z) on the basis of signals coming from in-core neutron flux detectors, temperature sensors, primary coolant flow and coolant pressure transducers. Calculated survey of possibility to use Kalman filters or corrective filter to eliminate time delay in SPND signals was carried out. The inaccuracy in the method of maximum linear power monitoring was determined. This work shows that the solution was found using the practice of in-core instrumentation, and ICIS designing and operation with improved metrology, reliability and time parameters and using advanced data communication technologies intended for important challenges of the world market, and as a response to standards.

Corrosion Phenomena of High Temperature Alloys by Helium Impurities in the Primary Coolant of HTGR

The primary coolant circuit of the high temperature gas-cooled reactor (HTGR) contains trace impurities. A nickel base alloy would corrode when exposed to an atmosphere at a high temperature and for a long time. The protective oxide scale formed by chromium is an important factor to prevent severe corrosion of high temperature alloys. Corrosion tests were conducted on Inconel 617, Incoloy 800H, Hastelloy X, and T-22, which are commonly used in the steam generator of HTGR. The alloys were exposed to helium with trace impurities for 48 hours at 950°C. The corrosion results were analyzed by weighing, scanning electron microscopy (SEM) and electron probe microanalyzer (EPMA). All the four alloys formed oxide scales in this atmosphere, but they differ in the capacity to resist corrosion. Therefore, the carbon transfer phenomenon observed in this experiment varies for the different alloys. In addition, for Cr in Inconel617, the expected depletion phenomenon near the corrosion layer occurred, which is consistent with the results from theoretical analysis.

High-resolution single-particle cryo-EM of samples vitrified in boiling nitrogen

Based on work by Dubochet and others in the 1980s and 1990s, samples for single-particle cryo-electron microscopy (cryo-EM) have been vitrified using ethane, propane or ethane/propane mixtures. These liquid cryogens have a large difference between their melting and boiling temperatures and so can absorb substantial heat without formation of an insulating vapor layer adjacent to a cooling sample. However, ethane and propane are flammable, they must be liquified in liquid nitrogen immediately before cryo-EM sample preparation, and cryocooled samples must be transferred to liquid nitrogen for storage, complicating workflows and increasing the chance of sample damage during handling. Experiments over the last 15 years have shown that cooling rates required to vitrify pure water are only ∼250 000 K s−1, at the low end of earlier estimates, and that the dominant factor that has limited cooling rates of small samples in liquid nitrogen is sample precooling in cold gas present above the liquid cryogen surface, not the Leidenfrost effect. Using an automated cryocooling instrument developed for cryocrystallography that combines high plunge speeds with efficient removal of cold gas, we show that single-particle cryo-EM samples on commercial grids can be routinely vitrified using only boiling nitrogen and obtain apoferritin datasets and refined structures with 2.65 Å resolution. The use of liquid nitrogen as the primary coolant may allow manual and automated workflows to be simplified and may reduce sample stresses that contribute to beam-induced motion.

Application of reverse osmosis at NPP and verification of the process for primary coolant treatment in temelín nuclear power

Drained primary coolant from nuclear power plants containing boric acid is currently treated in the system of evaporators and by ion exchangers. Reverse osmosis as an alternative process to evaporator was investigated. Using reverse osmosis, the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of concentrated boric acid solution together with other components, while permeate stream consists of purified water. In the first phase ofthe project the reverse osmosis modules from several manufactures were tested on a batch laboratory apparatus. Certain modifications to the pH of the feed solution were needed to enable the tested membranes to concentrate the H3BO3 in the retentate stream, separate from the pure water in the permeate stream. Furthermore, the separation capability for other compounds present in primary coolant such as K, Li or NH3 were evaluated. In the final phase of the project the pilot-plant unit of reverse osmosis was tested in nuclear power plant Temelín. It was installed in the Special Purification System SVO-6 for the regeneration of boric acid. The aim of the tests performed in Temelín nuclear power plant was to verify possible use of reverse osmosis for the treatment of primary coolant.

tudy on the determination of ¹⁰B/¹¹B isotope ratio in water samples by isotope dilution – inductively coupled plasma mass spectrometry (ID-ICPMS)

The determination of 10B/11B isotope ratio and boron concentration in various watersamples using isotope dilution technique with inductively coupled plasma mass spectrometry (ICPMS) was studied. The interferences on precision and accuracy in isotopic ratio determination by ICPMS such as memory effects, dead time, spectral overlap of 12C were investigated for the selection of optimum conditions. By the addition of certain amounts of enriched 10B into samples, the 10B/11B ratio was determined through ICP-MS signal of 10B and 11B. The detection limit for 10B and 11B was experimentally obtained as 0.26 µg/L and 0.92 µg/L, respectively. The ratios of 10B/11B in measured water samples varied in the ranged between 0.1905 and 0.2484 for different matrices. This method has been then applied for the determination of boron isotopic ratio in VVER-1000 reactor-type simulated primary coolant water and in some environmental water samples.

Investigation of Non-cooldown SG Secondary Condition on the Natural Circulation Cooling Procedure

In typical pressurized water reactor (PWR), in case that one steam generator (SG) cannot be credited for the primary cooldown, it is necessary to homogenize primary coolant temperature among loops using at least one reactor coolant pump (RCP) for the plant cooldown. If the natural circulation condition is established due to unavailability of all the RCPs, the continuous cooldown using intact SGs causes to disturb the smooth depressurization because it leads to void generation in the top of the non-cooldown SG tube where the high temperature coolant is remained. For this purpose, W.Sakuma, et al.[1] suggested the outline of asymmetric cooldown procedure without any RCPs restart. Since the suggested procedure is based on only one secondary condition (SG dry-out) of non-cooldown SG, and hence the impact of difference of the secondary condition should be investigated. In this paper, the sensitivity analyses were performed to confirm the impact on the asymmetric cooldown procedure, and consequently, it was confirmed that the coolable range used in the procedure was expanded if the water inventory exists in non-cooldown SG. Therefore it was concluded that the coolable range which was defined with the SG dry-out condition in non-cooldown SG can be conservatively applied for the operating procedure.