Loss Of Coolant Accident Analysis Research Articles

High-Speed Electrical Impedance Tomography of Fastly Evolving Two-Phase Flows for Loss of Coolant Accident Analysis

High-Speed Electrical Impedance Tomography of Fastly Evolving Two-Phase Flows for Loss of Coolant Accident Analysis

High-Speed Electrical Impedance Tomography of Fastly Evolving Two-Phase Flows for Loss of Coolant Accident Analysis

High-Speed Electrical Impedance Tomography of Fastly Evolving Two-Phase Flows for Loss of Coolant Accident Analysis

Development of drift–flux correlation for predicting void fraction in downcomer regions

ABSTRACT Accurate prediction of interfacial drag in the downcomer annulus is crucial for the assessment of downcomer void fraction for the loss of coolant accident analysis. The downcomer annulus is the gap between reactor pressure vessel (RPV) exterior and the inner wall of pressure containment vessel (PCV). Based on the previous research, occurrence of the nonuniform two-phase flow in downcomer section is reported, which is partly due to the large wall temperature difference between RPV exterior and the inner wall of PCV. In RELAP5, interfacial drag term in downcomer section is calculated using Kataoka–Ishii and churn-turbulent drift–flux correlations. It has been pointed out that this traditional calculation approach for calculating downcomer void fraction needs modification. The purpose of the current study is to assess the behaviors of drift–flux parameters in downcomer section and to propose an improved distribution parameter model that is suitable for donwcomer boiling analysis.

Loss of coolant accident analysis under restriction of reverse flow

Abstract This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

RISMC industry application #1 (ECCS/LOCA): Core characterization automation: Lattice codes interface for PHISICS/RELAP5-3D

The United States Nuclear Regulatory Commission (U.S. NRC) is currently proposing rule-making “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that the cladding performance cannot be evaluated anymore as separated from other disciplines, but needs to be evaluated in a coupled manner with other physics. The Risk-Informed Safety Margin Characterization (RISMC) Pathway initiated several programs at Idaho National Laboratory (INL) to support the industry in the transition to the new rule. One of these programs is the Industry Application #1 (IA1). IA1’s goal is to develop an Integrated Evaluation Model (IEM) called LOCA Toolkit for the US (LOTUS). This tool connects five major disciplines involved in LOCA analysis, among them is (automated) core design (CD-A). This paper focuses solely on the design of LOTUS' CD-A module as well as a demonstration application of the CD-A module to a generic Pressurized Water Reactor (PWR). The demonstration of the other LOTUS disciplines together with a demonstration example of the final risk-informed safety margin analysis is published in a separate paper. LOTUS is meant to provide the methodology framework into which the user can plug his established tools and codes. With the CD-A, the plant owner/operator will characterize his core design with the tools he has. He will then integrate his tools/methods into the LOTUS framework for the downstream safety analysis. For the current CD-A demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D, while cross section generation is done using the Studsvik lattice code HELIOS-2.

Evolution of licensing methodology for large break loss of coolant accident analysis and associated PCT margin quantification

The evaluation of methodology for large break loss of coolant accident (LBLOCA) licensing analysis involves two kinds of methodologies, namely deterministic methodology and risk-informed methodology. According to the existing 10 CFR50.46, LBLOCA is one of the most essential design-basis accidents and a deterministic methodology shall be applied to perform LBLOCA analysis based on a so-called surrogate sequence. Without considering how low this sequence occurrence probability is, this surrogate sequence satisfies all the required licensing assumptions. However, in the to-be-issued 10 CFR 50.46a, the LBLOCA can be categorized as an accident beyond design basis and the margin of peak cladding temperature (PCT) shall be evaluated in a risk-informed manner. Regarding the deterministic LBLOCA methodologies, the PCT margin quantified by the traditional appendix K methodology with conservative appendix K models and bounding plant parameters can be enlarged by 41.5K with the deterministic-realistic hybrid methodology (DRHM), in which realistic plant parameters and statistical uncertainty analysis are utilized. It was noted that the PCT margin released by realistic plant parameters can be about 1/4–1/5 of the margin generated by a full-scoped BEPU methodology with both best-estimate models and realistic plant parameters. Furthermore, by relaxing the licensing sequence assumptions, an additional LBLOCA PCT margin of 44.3–62.2K can be released from the deterministic approaches by applying the risk-informed safety margin characterization (RISMC) methodology. It was also observed from the risk-informed LBLOCA analysis, to have a cumulated occurrence probability over 99% in the load spectrum, the third sequence by order of ascending PCT was referred and the associated sequence occurrence probability of this sequence is 5.07*10−3. While in the deterministic methodology, the occurrence probability of the traditional surrogate or licensing sequence is only 5.46*10−5.

The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

Reduced order modeling of flashing two-phase jets

In the event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor, the escaping coolant produces a highly energetic flashing jet with the potential to damage surrounding structures. In LOCA analysis, the goal is often to evaluate many break scenarios in a Monte Carlo style simulation to evaluate the resilience of a reactor design. Therefore, in order to quickly predict the damage potential of flashing jets, it is of interest to develop a reduced order model that relates the damage potential of a jet to the pressure and temperature upstream of the break and the distance from the break to a given object upon which the jet is impinging. This work presents framework for producing a Reduced Order Model (ROM) that may be informed by measured data, Computational Fluid Dynamics (CFD) simulations, or a combination of both. The model is constructed by performing regression analysis on the pressure field data, allowing the impingement pressure to be quickly reconstructed for any given upstream thermodynamic condition within the range of input data. The model is applicable to both free and fully impinging two-phase flashing jets.

Small break LOCA analysis without emergency core cooling systems using the RELAP5/SCDAP code in VVER-1000 reactor

In this study, the loss of coolant accident analysis without emergency core cooling system on in-vessel core melt progression is evaluated under a severe accident scenario in Bushehr VVER-1000 reactor. Using the RELAP5/SCDAP mod3.4 computer code, a best estimate simulation event of a 25mm break SBLOCA, without ECCs is conducted until the reactor pressure vessel failure. In the performed analysis some important parameters such as the amount of hydrogen production, debris bed height and failure of lower plenum due to creep rupture are considered which are not included in the reactor FSAR.It has been concluded that, during the small breaks LOCA without emergency core cooling systems, the reactor core is uncovered at 7430s; consequently the initial relocation to the LP is occurred at 8130s. Based on the performed analysis the failure of the lower plenum is estimated in 9170s due to the creep rupture.

Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

During the loss of coolant accident (LOCA) of supercritical water cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2×2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior e.g. system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g. heat transfer coefficient, CHF correlation, are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

Loss of coolant accident analyses on multi purpose research reactor by RELAP5/MOD3.2 code

In this paper, a loss of coolant accident (LOCA) in multi purpose research reactor (MPRR) is analyzed. The following procedure consists of three major parts. First, a RELAP5 nodalization of the core and the most important components of the primarily cooling system are developed and steady state condition is simulated and compared with result of simulation a same Indian research reactor to verify the nodalization. Then, two large break points are selected for design basis accident analyses. Finally, two main loss of coolant accidents are simulated and developed to analyze LOCA in MPRR. It can be concluded that MPRR is safe against the LOCA.

The Plant Feature and Performance of Double MS (Modular Simplified and Medium Small Reactor)

A new concept of a small and medium sized light water reactor, named the double MS: modular simplified and medium small reactor (DMS) was developed. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts, and effective use of proven technology. The decrease in the primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about 2 m compared with 3.7 m in the conventional boiling water reactor (BWR). The short active fuel length reduces the drop in core pressure and overcomes the natural circulation system. By using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to confirm transient performance, the DMS plant performance under transient conditions was evaluated using the TRACG code. TRACG code, which can treat multidimensional hydrodynamic calculations in a RPV, is well suited for evaluating the DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As a result, the maximum change in the minimum critical power ratio of the DMS was 0.14, almost the same as for the current advanced boiling water reactor (ABWRs). In order to improve safety efficiency, developing an emergency core cooling system (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following: (1) eliminating high-pressure injection systems, (2) adopting passive safety-related systems, and (3) optimizing distribution for the systems and power source for the ECCS. In this way, the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of the loss of coolant accident analysis, we confirmed that the core can be covered by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.

Loss of coolant accident analyses on Tehran research reactor by RELAP5/MOD3.2 code

In this paper, a loss of coolant accident (LOCA) in Tehran research reactor (TRR) is analyzed. The following procedure consists of three major parts. First, a RELAP5 nodalization of the core and the most important components of the primarily cooling system are developed and steady state condition is simulated and compared with experimental data [SAR for TRR, October 2002. Safety analysis report for Tehran research reactor. Nuclear Research Center of the Atomic Energy Organization of Iran, Tehran] to validate the nodalization. Then, two large break points are selected for design basis accident analyses. Finally, two main loss of coolant accidents are simulated and developed to analyze LOCA in TRR. It can be concluded that TRR is safe against the LOCA. And also RELAP5/MOD3.2 code can simulate research reactors at operating conditions well but it seems that this code cannot simulates large transient phases of the research reactors accurately as steady state condition specially in low pressure two phase mixtures (water and air).

Influence of the break location in the loss of coolant accident analyses for the ITER divertor cooling loop

In the frame of the Generic Site Safety Report (GSSR) for the ITER experimental plant, several accident analyses have been carried out to analyse in detail the radiological risk linked with the possible releases. An ex-vessel loss of coolant followed by an in-vessel break, caused by the thermal stress on the divertor (DV) vertical target consequent to the plasma disruption, is the evolution of the accident referred in the study. The plasma disruption happens for the intervention of the fast plasma shutdown system occurring when the flow rate in the pump lowers below 80% of the nominal flow. This accident scenario is critical from the point of view of the vault pressurisation, of the activated materials mobilisation from the plasma chamber and of the hydrogen production inside the vacuum vessel (VV). The scope of the analyses presented in this paper is to quantify the influence of the ex-vessel break position on the accident consequences. A parametric study on the ex-vessel break position has been carried out. The pipe break before the heat exchanger is the most critical for the vault pressurisation and for the releases of activated corrosion products (ACP) and tungsten dust, while the break at the pump outlet maximises the hydrogen production inside the VV. The less critical accident is the break at the pump inlet for all the possible consequences.

Influence of Partial Blockage of a BWR Bundle on Heat Transfer, Cladding Temperature, and Quenching During Bottom Flooding or Top Spraying Under Simulated LOCA Conditions

In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m/sup 3//h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800/sup 0/C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis.

A Best-Judgment Analysis of Emergency Core Cooling System Performance

The requirements for an emergency core cooling system (ECCS) evaluation model that is acceptable for a pressurized water reactor licensing analysis are detailed in Appendix K to 10CFR50. The purpose of these requirements is to ensure that such an analysis will yield a conservative upper bound to the maximum cladding temperature and cladding oxidation that can result from a postulated loss-of-coolant accident (LOCA). By its nature, therefore, this model is inappropriate to indicate the actual anticipated results of a LOCA. Furthermore, a quantitative assessment of the conservatism inherent in the licensing model is unavailable. To produce realistic LOCA results, a calculation was performed at Combustion Engineering (C-E) for the reactor in its System 80/sup TM/ nuclear steam supply system, using a best-judgment ECCS evaluation model. The best-judgment model is a C-E first-generation best-estimate model that uses the basic Appendix K licensing computer programs, but in which the bounding conservatisms required by Appendix K are relaxed for selected parameters and models of primary concern in a LOCA analysis. The important differences between the best-judgment model and the Appendix K licensing model are examined.