Break Size Research Articles

Performance analysis of a multi-stage safety injection tank

A performance analysis of a multi-stage safety injection tank (MSIT) was carried out using a numerical method. A MSIT is capable of injecting coolant from a single tank in multiple stages in the event of a loss of coolant accident (LOCA) at a nuclear power plant. The performance and consequences of two- and three-stage MSITs were compared. The effects of break size, installation heights of the injection lines, and pressure-loss coefficients of each injection line were scrutinized. By selecting pertinent design parameters, the MSIT can satisfy its design requirements, by providing continuous coolant injection without problems of delays or overlaps for a minimum of 72 h after a LOCA.

Comparison of Two Different Sized Small-Break LOCAs on the Passive Safety Injection Line Using SMART-ITL Data

Even for small modular reactors (SMRs) with all large pipes removed, a small-break loss-of-coolant accident (SBLOCA) is an important design-basis accident (DBA). Experimental simulation of the SBLOCA scenario is essential before a prototype reactor is realized. The system-integrated modular advanced reactor (SMART) is one of the SMRs developed by the Korea Atomic Energy Research Institute. An integral test loop, SMART-ITL, was also constructed to carry out several types of integral thermal-hydraulic effects tests for the prototype reactor. The SMART-ITL was designed with a preserved height, 1/7th diameter, and 1/49th area, and volume-scaling ratios. Two types of passive safety systems were equipped in the SMART-ITL: a passive safety injection system (PSIS) and a passive residual heat removal system (PRHRS). The PSIS was designed to refill the coolant in the reactor coolant system (RCS) for 72 h after an accident. Under accident conditions the PRHRS prevents overheating and overpressurization of the RCS using two-phase natural circulation. The SBLOCA on the passive safety injection line is a significant DBA that should be validated for differences in break size. In this paper, the effects of two different break sizes, 2 and 7/32 in., were analyzed in order to study the effect of the maximum and minimum mass and energy loss of the RCS. In order to simulate a clear difference between maximum and minimum mass and energy loss of the RCS, heat removal by the PRHRS was performed in the maximum break size (2-in.) accident, and heat removal by the PRHRS was not conducted in the minimum break size (7/32-in.) accident. The difference in mass and energy loss of the RCS will have a significant impact on the operation of the automatic depressurization system. Using the two extreme accident simulations, it was possible to confirm the difference in accident progression caused by the difference in break size and the characteristics of the PSIS.

Continuous record Laplace-based inference about the break date in structural change models

Building upon the continuous record asymptotic framework recently introduced by Casini and Perron (2020a) for inference in structural change models, we propose a Laplace-based (Quasi-Bayes) procedure for the construction of the estimate and confidence set for the date of a structural change. It is defined by an integration rather than an optimization-based method. A transformation of the least-squares criterion function is evaluated in order to derive a proper distribution, referred to as the Quasi-posterior. For a given choice of a loss function, the Laplace-type estimator is the minimizer of the expected risk with the expectation taken under the Quasi-posterior. Besides providing an alternative estimate that is more precise—lower mean absolute error (MAE) and lower root-mean squared error (RMSE)—than the usual least-squares one, the Quasi-posterior distribution can be used to construct asymptotically valid inference using the concept of Highest Density Region. The resulting Laplace-based inferential procedure is shown to have lower MAE and RMSE, and the confidence sets strike a better balance between empirical coverage rates and average lengths of the confidence sets relative to traditional long-span methods, whether the break size is small or large.

Parameter importance ranking for SBLOCA of CPR1000 with moment-independent sensitivity analysis

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

NORMAL VALUES OF ESOPHAGEAL HIGH-RESOLUTION MANOMETRY: A BRAZILIAN MULTICENTER STUDY.

The high-resolution manometry has been a significant advance in esophageal diagnostics. There are different types of catheter and systems devices to capture esophageal pressures that generate variable data related to Chicago Classification (CC) and consequently influence normal values results. There are not normative data for the 24-channel water-perfused high-resolution manometry system most used in Brazil with healthy volunteers in supine posture. To determine manometric esophageal normative values for a 24-channel water-perfused high-resolution manometry catheter in supine posture using healthy volunteers according to CC 3.0 parameters. A total of 92 volunteers with no gastrointestinal symptoms or medications affecting GI motility underwent esophageal high-resolution manometry by standard protocol. Age, gender and manometry parameters analyzed using Alacer software were collected. The median, range, and 5th and 95th percentiles (where applicable) were obtained for all high-resolution manometry metrics. Normal value percentiles were defined as 95th integrated relaxation pressure, 5th-100th distal contractile integral, and 5th distal latency. The mean age was 40.5±13.2 years. Our normative metrics were integrated relaxation pressure <16 mmHg and distal contractile integral (708-4111 mmHg.cm.s) distal latency was <6 s and peristaltic break size (>4 cm). For EGJ-CI the range 5th-95th was 21.7-86.9 mmHg.cm.s. This is the first report of normative data for the 24-channel water-perfused system in supine posture. It revealed higher integrated relaxation pressure and distal latency duration which suggest the need to change CC 3.0 cutoffs for this system. It is observed that there is a tendency that DCI >7000 mmHg.cm.s may represent the lower limit of hypercontractility, and when <700 mmHg.cm.s (<5% percentile) interpreted as ineffective esophageal motility or failcontraction. Also compared to Chicago 3.0, higher integrated relaxation pressure and duration of distal latency were found. We emphasize that these data must be confirmed by future studies.

Core makeup tank behavior investigation during ACME integral effect tests

Passive safety systems are widely used in the advance reactors, such as AP1000 and CAP1400. As an important component in the passive safety system, the core makeup tank (CMT) plays a key role for the safety injection and the core decay heat removal. Since the CMT not only provides the full-pressure injection but also triggers the auto depressurization system by its level signal during small-break loss of coolant accident (SBLOCA), the CMT thermal hydraulic behaviors affect the SBLOCA transient process. The thermal hydraulic behaviors of the CMT were investigated by conducting different SBLOCA tests on the ACME integral effect test facility, which includes the 2.5 cm, 5 cm and 20 cm breaks, and double ended direct vessel injection (DVI) line break. Based on the test observations, the CMT injection response to SBLOCA was divided into four phases: the circulation, the drain transition, the injection inhabitation and the final drain. The CMT injection characteristics and the dominate factors were scrutinized based on the phase division. It was found that the break condition has significant impacts on the CMT operating characteristics in each phase. The CMT circulation flow rate can be predicted, and the break size determines the circulation phase length. The CMT drain transition process is controlled by the PBL voiding process that has a close relation to the liquid level change in the cold leg. The ACC-CMT interaction was modeled, and the prediction result agreed with the test result. The comparison tests showed that the ACC nitrogen gas injection still affected the CMT drain behavior. A similar thermal stratification pattern formed during the different SBLCOA tests, and the thermal stratification remained not only in the CMT internal fluid but also in the CMT wall. The estimation of the flashing amount showed it did not play a significant role for the CMT liquid level decline, but it could induce the increases in the pressure difference between the CMT and the system and thus accelerate the CMT drain rate. The CMT internal fluid temperature determined the wall temperature, and a reverse heat transfer process occurred due to the flashing induced wall cooling during the ADS depressurization.

High Homology-Directed Repair Using Mitosis Phase and Nucleus Localizing Signal.

In homology-directed repair, mediated knock-in single-stranded oligodeoxynucleotides (ssODNs) can be used as a homologous template and present high efficiency, but there is still a need to improve efficiency. Previous studies have mainly focused on controlling double-stranded break size, ssODN stability, and the DNA repair cycle. Nevertheless, there is a lack of research on the correlation between the cell cycle and single-strand template repair (SSTR) efficiency. Here, we investigated the relationship between cell cycle and SSTR efficiency. We found higher SSTR efficiency during mitosis, especially in the metaphase and anaphase. A Cas9 protein with a nuclear localization signal (NLS) readily migrated to the nucleus; however, the nuclear envelope inhibited the nuclear import of many nucleotide templates. This seemed to result in non-homologous end joining (NHEJ) before the arrival of the homologous template. Thus, we assessed whether NLS-tagged ssODNs and free NLS peptides could circumvent problems posed by the nuclear envelope. NLS-tagging ssODNs enhanced SSTR and indel efficiency by 4-fold compared to the control. Our results suggest the following: (1) mitosis is the optimal phase for SSTR, (2) the donor template needs to be delivered to the nucleus before nuclease delivery, and (3) NLS-tagging ssODNs improve SSTR efficiency, especially high in mitosis.

EDF MAAP5.04 LOCA Transient Simulation Improvements

During a loss-of-coolant-accident (LOCA) transient in a pressurized water reactor (PWR), water from the primary circuit is lost at the break. PWR designs are equipped with safety systems (SS) such as safety injection or accumulators to inject water into the primary circuit and prevent the core from being degraded. Depending on the size, position, and orientation of the break, a part of the safety system injection (SSI) into the primary circuit will be lost at the break. This parameter has a significant influence on the time the core uncovers in case the SS are lost. MAAP5.04 enables users to define the part of SSI that is lost at the break. Apart from a double-ended–break LOCA transient, users struggle to define precisely the part of SSI lost at the break, but this choice can have an important impact on the transient key event times. Thanks to its detailed equations and nodalization, the reference Code for Analysis of Thermal Hydraulics during an Accident of Reactor and safety Evaluation (CATHARE) enables one to evaluate the part of SSI lost at the break. Numerous CATHARE calculations have been performed taking into account different break sizes, positions, and orientations to determine the part of SSI lost at the break in each case. A metamodel has been created from the constituted database and implemented in EDF MAAP5.04. This paper also presents the impact of these improvements on LOCA transients where SS are lost.

Preliminary optimization of emergency core cooling system with accident-tolerant fuels

Based on the excellent performance of accident-tolerant fuels (ATFs), the optimization schemes for the safety system of Chinese advanced three-loop pressurized water reactor during loss of coolant accident (LOCA) were proposed and corresponding feasibility analysis was carried out in this study. The modified system analysis code was used to safety evaluation and four ATF systems were investigated. Several modifications were made to the rod geometries in partial ATF systems to make up the poor neutron economic performance of FeCrAl and MoMAX. The emergency core cooling system (ECCS) was optimized during LOCA with different break sizes. The results show that it is feasible for all ATF systems to cancel the high pressure safety injection (HPSI) or cancel both HPSI and low pressure safety injection (LPSI) but add an extra accumulator (ACC). Also the safety of reactors fueled with ATFs can be guaranteed with only 0.35 times the original flowrate of LPSI during large break LOCA (LB-LOCA). In the scheme of replacing the original ECCS with two extra ACCs during LB-LOCA, reactors loaded with partial ATFs are safe, while UO2-Zr systems are in danger as a contrast. This work could provide reference for the design of reactor safety system in the future.

Designing a model for predicting cyber-attack breaches in an untrusted environment by using linear regressing algorithm

Breaking down digital episode informational indexes is a significant technique for developing our comprehension of the advancement of the danger circumstance. This is a generally new research point, and numerous investigations stay to be finished. In this paper, we report a measurable investigation of a rupture occurrence informational index comparing to 12 years (2005–2017) of digital hacking exercises that incorporate malware assaults. We show that, as opposed to the discoveries revealed in the writing, both hacking rupture occurrence between appearance times and break sizes ought to be demonstrated by stochastic procedures, instead of by appropriations since they display autocorrelations. At that point, we propose specific stochastic procedure models to, separately, fit the between appearance times and the rupture sizes. We additionally show that these models can anticipate the between appearance times and the break sizes. So as to get further bits of knowledge into the advancement of hacking break episodes, we direct both subjective and quantitative pattern examinations on the informational index. We draw a lot of digital security bits of knowledge, including that the danger of digital hacks is without a doubt deteriorating regarding their recurrence, however not as far as the greatness of their harm.

Best estimate plus uncertainty analysis of small break loss of coolant accidents with different break sizes for CPR1000

The best estimate plus uncertainty (BEPU) is an advanced method which has been widely adopted in nuclear power plant safety analysis. In order to investigate the transient behaviors of the pressurized water reactor CPR1000 during the cold leg small break loss of coolant accident (SBLOCA) at different break sizes, BEPU analysis of the SBLOCA was carried out in this study, and the sizes of the break were 0.2%, 0.4%, 0.6%, 0.8% and 1.0% of the cross-sectional area of the cold leg, respectively. The minimum collapsed level (MCL) in the active core section was determined to be the figure of merit (FOM). The base case calculations, uncertainty propagation calculations and sensitivity analysis calculations were performed with the NUSOL-SYS code. Based on the results, the transient behaviors were analyzed, the lower 95/95 tolerance limits of the MCL at different break sizes were quantified, and the important input parameters were determined.

Investigation of a best oxidation model and thermal margin analysis at high temperature under design extension conditions using SPACE

Abstract Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T 1853 K) were in excellent agreement with the data of the QUENCH experiments. For ‘LOCA without SI’ (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the ‘LBLOCA without SI’ and ‘SBLOCA without SI’ accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

Re-assessment of accumulators performance to identify VVER-1000 vulnerabilities against various break sizes of SB-LOCA along with SBO

After Fukushima Daiichi accident it is suggested to identify the vulnerabilities of nuclear power plants. In SB-LOCA along with SBO scenario, the passive safety systems can have a positive influence on the accident mitigation. This article aims to analyze the performance of passive accumulators (ACC) injection in the spectrum (10–100 mm break sizes) of SB-LOCA that simultaneously occurs with SBO in Bushehr Nuclear Power Plant (BNPP). The results showed that ACCs are not able to inject in the whole range of breaks, because the primary loop pressure is not decreased sufficiently in minimal breaks without operator intervention. Also, the Core Damage Times (CDT) have been calculated for this range of breaks based on 1204 °C of Peak Cladding Temperature (PCT) and Burst Pressure-Temperature Diagram (BPTD) to identify the vulnerable break size. Further, analysis showed that the range of smaller breaks has a shorter CDT (the 50 mm break is the most vulnerable break size) that can be improved by optimizing the set-points of accumulators.

Comparison of loop seal clearing behavior between SBLOCA and SBO-related tests in the ATLAS facility

Loop seal clearing (LSC) behavior was investigated for small-break loss-of-coolant accident (SBLOCA) and station blackout (SBO)-related tests in the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility. Major characteristics of LSC behavior between two scenarios were identified and summarized. In SBLOCA tests, the first LSC provided an instant event connecting the pressure plateau phase to the boil-off phase, and tended to induce a resultant peak cladding temperature (PCT) for larger break sizes. Every LSC induced an intermittent inventory redistribution between the downcomer (DC) and the core. In SBO-related tests, most LSCs occurred under a nearly stagnant or even reverse loop flow condition, especially in the hot legs, and showed little effect on PCT occurrence. After LSC, water level changes in the core and loop seals showed synchronizing trends but opposite directions. LSC also induced mass transfer between the DC and core. And a visual observation on the sequence of an LSC was discussed.

Monte Carlo Evaluation of the Methods Estimating Structural Change Point in Panel Data

In this study, we investigate the existence of structural break in a panel data consisting of N time series of T unit length, and the estimation performance of Simple Mean Shift Model, Fluctuation Test, Wald Statistic Test, Kim Test which are based on common break assumption are examined to determine the break date. In this context, 108 Monte Carlo simulations are performed, each of which consisted of 3000 repetitions for the factors number of cross-sections, time dimension, break size and break rate factors, which are considered to influence the performance of the tests. As a result of the Monte Carlo simulations, the Simple Mean Shift Model approach predicts the break point with a higher performance than the other methods. In addition, if the breakpoints are at the midpoint of the series, the Wald Statistic and Kim Tests show the highest performances, while the Fluctuation Test shows the highest breakpoint predictive performance if break occur in the third quarter of the series. Generally, as the number of cross-sections increases, the estimation performance of the tests increases, whereas as the time dimension increases, the performance of methods other than the Simple Mean Shift Model decreases. As a final point, it has been observed that there is no significant effect of the break size on the predictive performance of the methods.

Assessment of VVER 1000 core degradation for bounding cases with ASTEC 2.1.1.0

This paper presents an assessment and comparison of the core degradation progression in four Station Blackout (SBO) scenarios analysed by the ASTEC computer code. Two types of SBO scenarios namely with high- and low-pressure conditions are considered. The low-pressure conditions in the SBO scenarios were simulated by introducing a small break loss of coolant accident in the cold leg with an equivalent internal diameter of 80 mm. The selection of the break size was based on the idea to have a significantly faster primary side pressure reduction for simulation of low-pressure conditions and a significantly longer coolant injection from the hydro accumulators (passive safety system).For both types of selected scenarios, the plant behaviour was analysed without operator actions, which allows assessing the times to reach the important set points during the accident progression. The set points include; a) the prediction of the dryout of the steam generators (SG) (or loss of SG effectiveness), b) the loss of natural circulation leading to core uncovery and heat up, c) the beginning of hydrogen generation, d) different stages in core degradation, e) the actuation of the passive safety system etc. The analyses were performed until reactor vessel failure takes place in both of the investigated scenarios. The purpose of these analyses is to study the reactor core behaviour parameters and to estimate the time available to perform operator actions.In addition, this investigation is focused on the assessment of the effectiveness of operator actions as prescribed in the Severe Accident Management Guidlines (SAMGs) for the investigated reactor type. The referenced NPP is KNPP equipped with two VVER 1000 V320 reactors. The ASTECv2.1.1.0 computer code has been used for the investigation. The aim of these analyses is to assess the possibility preserving the reactor core from damage during a severe accident and to assess the hydrogen generation that occurs as a result of the overheated core reflooding at high- and low-pressure initial conditions. The injection of a coolant by an active system will start at the same core exit temperature for each scenario, but at different pressure.

Determination of pressure loss coefficient of the hot duct break of the CEA ALLEGRO 2009 concept using a CFD technique

ALLEGRO is a demonstration reactor of the GEN IV GFR2400 gas fast reactor concept which is currently in the pre-conceptual design phase. In the present ALLEGRO design, the hot duct (leg) – which connects the reactor with the main heat exchanger – is located inside the cold duct. For this reason, the break of the hot duct does not lead to a classical loss of coolant accident (there is no depressurization of the system), but a core by-pass occurs, which may threaten the integrity of the core. The peak cladding temperature (PCT) highly depends on the pressure loss between the hot and the cold ducts at the break. In this study, by an original approach we aimed to estimate the value of the pressure loss coefficient between the hot and the cold ducts by using the ANSYS FLUENT CFD (Computational Fluid Dynamics) code as a numerical tool. The results showed that if the break is located in the hot duct halfway between the reactor vessel and the main heat exchanger, the pressure loss coefficient value falls in the range of 1.05–1.15 for break sizes up to 100%. Considering the primary importance of the hot duct break pressure loss coefficient, further studies are proposed to assess the worst case break geometry.

Numerical analysis of loss of vacuum accident (LOVA) and preliminary discussion about dust resuspension for CFETR

Loss of vacuum accident (LOVA) is the air ingress into the vacuum vessel (VV) due to the failure of components. The high speed air jet flow induced by the pressure difference between the ambient condition and the in-vessel condition may cause the dust migration, re-suspension bring risk to the integrity of components and radioactive release. The present work deals with the computational fluid dynamics (CFD) modeling of airflow characteristics in different scenarios during the LOVA for CFETR. Then, forces applied on the deposited particles are investigated, based on which the preliminary discussion are carried out to analyze the possibility of dust resuspension in various LOVA scenarios. The results show that the break location, break size and the operational status during the LOVA have significant influence on the key parameters related to airflow forces such as the friction velocity. Furthermore, the airflow force gradually plays the dominate role as the particle diameter increases. It preliminarily demonstrates the possibility of dust resuspension as the particle diameter changing. The results can provide references to study the dust re-suspension and implement accident mitigation measures.

Experimental Determination of the Fire Break Size and Specific Water Consumption for Effective Control and Complete Suppression of the Front Propagation of a Typical Ground Fire

Experimental Determination of the Fire Break Size and Specific Water Consumption for Effective Control and Complete Suppression of the Front Propagation of a Typical Ground Fire

Real-time estimation of break sizes during LOCA in nuclear power plants using NARX neural network

Abstract This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms. NARX neural network is able to directly deal with time-dependent signals for dynamic estimation of break sizes in real-time. The case studied is a LOCA in the primary system of Bushehr nuclear power plant (NPP). In this study, number of hidden layers, neurons, feedbacks, inputs, and training duration of transients are selected by performing parametric studies to determine the network architecture with minimum error. The developed NARX neural network is trained by error back propagation algorithm with different break sizes, covering 5%–100% of main coolant pipeline area. This database of LOCA scenarios is developed using RELAP5 thermal-hydraulic code. The results are satisfactory and indicate feasibility of implementing NARX neural network for break size estimation in NPPs. It is able to find a general solution for break size estimation problem in real-time, using a limited number of training data sets. This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr NPP.