• Nuclear Engineering and Technology
• /
• v.45 no.7
• /
• pp.895-910
• /
• 2013

The reflood test data from the rod bundle heat transfer (RBHT) test facility showed that the grids in the upper portion of the rod bundle could become wet well before the arrival of the quench front and that the sizes of liquid droplets downstream of a wet grid could not be predicted by the droplet breakup models for a dry grid. To investigate the water droplet generation from a wet grid spacer, a viscous linear temporal instability model of the water sheet issuing from the trailing edge of the grid with the surrounding steam up-flow is developed in this study. The Orr-Sommerfeld equations along with appropriate boundary conditions for the flow are solved using Chebyshev series expansions and the Tau-Galerkin projection method. The effects of several physical parameters on the water sheet oscillation are studied by determining the variation of the temporal growth rate with the wavenumber. It is found that a larger relative steam velocity to water velocity has a tendency to destabilize the water sheet with increased dynamic pressure. On the other hand, a larger ratio of steam boundary layer to the half water sheet thickness has a stabilizing effect on the water sheet oscillation. Droplet diameters downstream of the spacer grid predicted by the present model are found to compare reasonably well with the data obtained at the RBHT test facility as well as with other data recently reported in the literature.

• Nuclear Engineering and Technology
• /
• v.12 no.3
• /
• pp.153-162
• /
• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.

• Nuclear Engineering and Technology
• /
• v.42 no.6
• /
• pp.690-703
• /
• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1532-1539
• /
• 2019

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.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05b
• /
• pp.429-434
• /
• 1996

K-REM［1］, which is under development as a realistic evaluation model of large break LOCA, is applied to the analysis of cold leg guillotine break of Ulchin 3&4. Fuel parameters on which statistical analysis of their effects on the peak cladding temperature (PCT) are made and system parameters on which the concept of limiting value approach (LVA) are applied, are determined from the single parameter sensitivity study. 3 parameters of fuel gap conductance, fuel thermal conductivity and power peaking factor are selected as fuel related ones and 4 parameters of axial power shape, reactor power, decay heat and the gas pressure of safety injection tank (SIT) are selected as plant system related ones. Response surface of PCT is generated from the plant calculation results and on which Monte Carlo sampling is made to get plant application uncertainty which is statistically combined with code uncertainty to produce the 95th percentile PCT. From the break spectrum analysis, blowdown PCT of 1350.23 K and reflood PCT of 1195.56 K are obtained for break discharge coefficients of 0.8 and 0.5, respectively.

• Nuclear Engineering and Technology
• /
• v.26 no.3
• /
• pp.355-366
• /
• 1994

The USNRC issued a revised ECCS rule that allows the use of best estimate computer codes for safety analysis. The rule also requires an estimation of uncertainty in calculated system response when applying the best estimate computer codes. A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the ECCS rule has been developed and this paper describes the application of new realistic evaluation methodology to large break LOCA for, the demonstration of the new methodology. The computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/MOD3.1, was used as the best estimate code in the application. The uncertainty of the code was evaluated by assessing several separate and integral effect tests, and for the application to actual plant Kori 3 & 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by random sampling or Monte-Carlo method for each response surface. Final uncertainties were quantified at 95% probability level and safety margins for large break LOCA were discussed.

• Nuclear Engineering and Technology
• /
• v.44 no.7
• /
• pp.727-734
• /
• 2012

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.

• Journal of Energy Engineering
• /
• v.14 no.1
• /
• pp.54-61
• /
• 2005

The APR1400 is an Advanced Pressurized Water Reactor with 3983 MWt power, 2×4 loops, and direct vessel injection system. The Fluidic Device (FD) is adopted to regulate the safety injection flow rate in a Safety Injection Tank (SIT) of APR1400. The performance of a newly designed fluidic Device is evaluated by analyzing a Large Break Loss-of-Coolant Accident (LBLOCA) using TRAC-M/F90, version 3.782. The analysis results show that the TRAC-M code reasonably predicts the important phenomena of blowdown, refill and reflood phases of LBLOCA. The sensitivity studies about gas/water volume changes in a SIT and K factor changes in a SI system were also done to understand the important phenomena with a Fluidic Device in APR1400.

• Nuclear Engineering and Technology
• /
• v.23 no.1
• /
• pp.56-65
• /
• 1991

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.

• Nuclear Engineering and Technology
• /
• v.21 no.4
• /
• pp.259-266
• /
• 1989

An improved version of RELAP5/MOD2 Cycle 36.04 code is assessed for LOFI LBLOCA Test L2-5. Minor modifications to the original version have been done to avoic reflood related errors. Based on the modified version, one base case and two cases for sensitivity study on downcomer and core channel modelling are calculated. The calculation results are compared with the experimental data for primary system pressure, break mass How rate and cladding temperature at hot spot According to the comparison, it is found that the hydraulic system behaviors are well predicted, excessive core cooling exist in blowdown phase for a single core channel and a combined downcomer case, and a better result can be obtained for a two core channel case.