Emergency Core Cooling System Performance Research Articles

CFD simulation of injection point design for emergency core cooling system of ALLEGRO

The Emergency Core Cooling System (ECCS) is one of the critical safety systems in the Gas-cooled Fast Reactor (GFR) demonstrator ALLEGRO. The ECCS design must be improved to enhance fundamental safety functions, particularly in removing heat from the core under postulated plant conditions. In this paper, we focused on exploring the effect of coolant injection point on the performance of ECCS, for more efficient core cooling under Loss Of Coolant Accident (LOCA) scenarios. We initially performed a 1D primary system loop thermal hydraulic simulation to determine appropriate boundary conditions in case of a LOCA. These conditions were then applied to a CFD model of the ALLEGRO reactor pressure vessel, including injection nozzles at two positions. Transient 3D CFD simulations of LOCA in ALLEGRO were carried out using Fluent to determine location of the most effective emergency coolant injection point. We simulated the flow distribution through the core up to around 30 s after both small- and large-break LOCA events. The results suggest that bottom injection of the emergency coolant is preferred in an updated ECCS design. It provides a more uniform flow distribution across the core and therefore more advantageous than the previously suggested location in the downcomer.

Emergency core cooling system performance criteria for Multi-Layered Silicon Carbide nuclear fuel cladding

Multi-layered Silicon Carbide (SiC) cladding has been proposed to replace the current Zircaloy (Zr) cladding in existing nuclear power plants. Due to its ceramic property, SiC may have a different failure mechanism from that of metallic Zr clad. This characteristic should be considered in designing the plant’s safety system performance, such that the fuel’s integrity is reasonably assured during accident scenarios.This paper aims at optimizing the Emergency Core Cooling System (ECCS) performance criteria in mitigating LBLOCA accidents at a Pressurized Water Reactor (PWR) using the uranium dioxide fuel and duplex or triplex SiC cladding structures. Clad failures by brittle fractures and thermal degradation were modeled in a fuel analysis code. This fuel model was coupled with the RELAP5 plant model by adapting the fuel’s mesh configuration and formulating the effective heat transfer coefficients. A model to quantify the likelihood of clad functional failure as a combination of individual layer fractures at various axial positions was proposed. Updated SiC material properties were gathered from recent reference studies. Clad reliability estimates using the proposed model were compared with the estimates obtained using other models. The results were used to relax the criterion of required actuation timing of ECCS, such that the risk of the new plant design is less or equal to that of the initial plant configuration.Triplex clad was more reliable than duplex clad and Zr, reducing the conditional core damage probability by three orders of magnitude in the reference LBLOCA scenario. Compressive stresses due to irradiation swelling played the determining role in the vulnerabilities of each clad design. The duplex clad was vulnerable to pressure-loading stresses while triplex clad was prone to fail due to thermal shocks experienced during the LOCA reflood phase. Additionally, it was concluded that the most vulnerable axial fuel rod location in the reference LBLOCA scenario was located at the highest axial power peaking location.A sensitivity study on the ECCS timing density was conducted. The results showed that ECCS actuation timing could be extended from the initial technical specification, without increasing the fuel damage likelihood. The time extension was distributed in the order of several hundreds of seconds, which was comparable with time extensions offered by FeCrAl and Cr-coated-Zr clad designs in LBLOCA scenarios. This time extension may potentially reduce the failure probabilities of EDGs and pumps, hence improving the plant safety further.

Modeling of flow blockage due to swell and rupture for analysis of LBLOCA using a multiple fuel rods scheme under high burnup condition

The present study discusses a model of Large Break Loss-Of-Coolant Accident (LBLOCA) calculation using the multiple fuel rods modeling scheme which was developed to consider the fuel rods with different burnup in preparation for the recently proposed revision of Emergency Core Cooling System (ECCS) rule in the United States of America and in Korea. The thermal conductivity of the fuel pellet, the initial oxide layer thickness of the cladding, the size of gap and the rod internal pressure, and the effective thermal conductivity of the cladding considering the oxide layer are reflected by an input as functions of the burnup. Further, through the analysis on the division of hydraulic channels in which the multiple fuel rods are located, the influences of division of the hydraulic channel and the crossflow between channels on cladding thermal response are evaluated. Also a modeling scheme to calculate the flow blockage due to swell and rupture of cladding for the current system thermal-hydraulic code such as MARS-KS code is discussed. To consider the uncertainty of the swell and rupture model of the code and the uncertainty in the present flow blockage modeling scheme, a multiplication factor is introduced and used to evaluate the influence of the flow blockage on ECCS performance.

10 CFR 50.46c Rulemaking: A Novel Approach in Restating the LOCA Problem for PWRs

The U.S. Nuclear Regulatory Commission (NRC) is considering a rulemaking that would revise requirements in 10 CFR 50.46 [also known as the emergency core cooling system (ECCS) rule]. Experimental work sponsored by the NRC suggested that the current regulatory acceptance criteria on ECCS performance during design-basis accidents are actually nonconservative for higher-burnup fuel, that embrittlement mechanisms not contemplated in the original criteria exist, and that the 17% limit on oxidation is not adequate to preserve the level of ductility that the NRC originally deemed to be warranted for adequate protection. The new rule imposes new acceptance criteria and is expected to be in effect within this decade. An implementation plan was developed that will give individual plants up to 7 years with which to comply once the rule is amended, depending on the status of each plant's analysis of record, the effort involved, and existing analytical margin to the limits.The proposed rule may challenge U.S. light water reactor fleet operational flexibility and economics. Within the U.S. Department of Energy Light Water Reactor Sustainability Program, the Idaho National Laboratory is pursuing an initiative that is focused on industry applications using Risk-Informed Safety Margin Characterization (RISMC) tools and methods applied to issues that are of current interest to the operating fleet. The mission of RISMC is to provide cost-beneficial approaches to safety analysis by leveraging modern methods, augmented tools (a combination of existing and new), and repurposed data (existing, but used in a new way).

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology

The AREVA NP Inc. realistic large-break loss-of-coolant-accident (LOCA) analysis methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. This methodology conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the United States Nuclear Regulatory Commission in the late 1980s. In addition, several practical considerations were revealed with the move to a production application. This paper describes the methodology development within the CSAU framework and utility objectives, lessons learned, and insight about current LOCA issues.

Regulatory Scenario for the Acceptance of Uncertainty Analysis Methodologies for the LB-LOCA and the Brazilian Approach

The task of regulatory body staff reviewing and assessing a realistic large break loss-of-coolant accident evaluation model is discussed, facing the actual regulatory licensing environment related to the acceptance of the analysis of emergency core cooling system performance. Especially, focus is directed to the question of how to fulfill the requirement of quantifying the uncertainty in the calculated results when they are compared to the acceptance criteria for this system. As it is recognized that the regulation governing the loss-of-coolant accident analyses was originally developed by the United States Nuclear Regulatory Commission, a description of its evolution is presented. When using a realistic evaluation model to analyze the loss-of-coolant accident, different approaches have been used in the licensing arena. The Brazilian regulatory body has concluded that, in the current environment, the independent regulatory calculation is recognized as a relevant support for the staff decision within the licensing framework of a realistic analysis.

The State-of-the-Art Theory and Applications of Best-Estimate Plus Uncertainty Methods

The approval of the revised rule on the acceptance of emergency core cooling system performance in 1988 triggered a significant interest in the development of codes and methodologies for uncertainty evaluation of best-estimate loss-of-coolant accident (LOCA) analyses. The code scaling, applicability, and uncertainty evaluation method was developed and demonstrated for a large-break LOCA in a pressurized water reactor. Later, several new best-estimate plus uncertainty methods (BEPUs) were developed around the world. The purpose of this paper is to identify and compare the statistical approaches of BEPU methods and present their importance for licensing applications in nuclear power plants. The study showed that the uncertainty analysis with random sampling of input parameters, using the nonparametric statistical tolerance limits for estimating uncertainty of output parameters, is the commonly accepted approach today. The existing BEPU methods seem mature enough, while the future research may be focused on the codes with internal assessment of uncertainty.

AREVA's realistic large break LOCA analysis methodology

The AREVA realistic large break loss-of-coolant-accident (LOCA) analysis methodology received approval by the United States Nuclear Regulatory Commission (USNRC) in April 2003. The methodology references the 1988 amended 10 CFR 50.46 allowing best-estimate calculations of emergency core cooling system performance. In addition, it conforms to the code scaling, applicability, and uncertainty (CSAU) methodology developed by the Technical Program Group for the USNRC in the late 1980s; however, it also incorporates the nonparametric statistical approach originally incorporated in the Gesellschaft fur Anlagen und Reaktorsicherheit (GRS) methodology for LOCA analysis. This paper provides a methodology summary, lessons learned and insight about current LOCA issues.

Transport Characteristics of Selected Pressurized Water Reactor LOCA-Generated Debris

In the unlikely event of a loss-of-coolant accident (LOCA) in a pressurized water reactor, break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS.A systematic study was conducted on various types of fibrous and metallic foil debris to determine their transport in water. Test results reported include incipient movement, bulk movement, accumulation on a screen, the ability of debris to jump over 5-cm (2-in.) and 15-cm (6-in.) curbs, and the effects of accelerating flow and turbulence. These data are currently being used in conjunction with computational fluid dynamics modeling to determine the potential for each debris type to reach the suction screen.

Design options for the safety injection system of Korean next generation reactor

Loss of coolant accident (LOCA) analyses for various configurations of safety injection system (SIS) are performed to optimize the emergency core cooling system (ECCS) performance for the Korean next generation reactor (KNGR). The KNGR is an advanced light water reactor (ALWR) adopting the advanced design feature of a direct vessel injection (DVI) configuration and passive fluidic device in the discharge line of the safety injection tank (SIT). To determine the feasible SIS configuration and the optimum capacities of the SIT and high pressure safety injection pump (HPSIP), licensing design basis and best estimate LOCA analyses are performed for the limiting large break and small break spectrum, respectively. The analyses results show that the four-train DVI injection with the current system design is a more feasible configuration than the other ones considered and the adoption of a fluidic device SIT enhances the ECCS performance for large break LOCA. For small break LOCA, in the case of cold leg break, the DVI4 configuration is better than other configurations and also meets the EPRI ALWR requirement of no core uncovery for up to a 15.24 cm (6 in) diameter small break. However, in the case of DVI line break, slight core uncovery is predicted and also the system behavior is significantly affected by reactor vessel (RV) downcomer modeling. Therefore, the DVI4 configuration is more feasible for KNGR ECCS performance, but further investigations are required to resolve the ECCS bypass issues for large break LOCA and to develop a proper RV downcomer model for analysis of DVI line break in small break LOCA.

Quantitative evaluation of ECCS capability during refill-reflood phase for BWR LOCA.

In this study, to evaluate the core cooling characteristics and water level transient more realistically during refill-reflood phase for a BWR LOCA, several separate effects tests have been carried out. Several correlations and physical models, based on this study, were utilized to develop a computer code SAFER03. The verification of SAFER03 by ROS Affi and FIST- ABWR tests showed that the SAFER03 predicted well the water level transient and the peak cladding temperature with some margin. This study enabled a realistic prediction of BWR emergency core cooling system performance during LOCA. This paper is a summary report on the study results.

Loss-of-Coolant Accident Performance of the Westinghouse 600-MW(electric) Advanced Pressurized Water Reactor

Emergency core cooling system performance analyses that demonstrate compliance with the 10CFR50.46 final acceptance criteria have been performed for the unique Westinghouse AP600 advanced pressurized water reactor. The AP600 plant features a low power density core, canned-motor reactor coolant pumps, passive safeguards systems, and an automatic depressurization system. It exhibits exceptional loss-of-coolant accident (LOCA) performance

Determination of the Net Free Containment Volume at the Trojan Nuclear Plant Using an Experimental-Analytical Technique

The net free volume of the containment is an essential parameter in the loss-of-coolant accident (LOCA) containment pressure analysis for pressurized water reactors. For an optimized emergency core cooling system performance due to the importance of backpressure during the reflood phase of a LOCA, it is necessary to have the predicted pressure quite close to the design pressure. Using a geometric analysis, an estimate of 56 241.99 m3 (1 986 167 ft3) for the net free volume has been made for the containment of the Trojan nuclear plant.Two sets of data were produced from the normally scheduled structural integrity and integrated leak-rate tests on the Trojan containment. These data sets were used to arrive at two new estimates of the net free volume. A deterministic equation giving volume as a function of the slope of a linear relationship between depressurization and time was developed. After an analysis of the reliability of the data, estimates of this linear slope were made from the two data sets. These two slopes gave net free volume estimates of 58 000 m3 (2.05 × 106 ft3) and 57 650 m3 (2.036 × 106 ft3) when used in the deterministic equation. The maximum deviation from the geometric-based estimate was <4%.