Emergency Core Cooling Injection Research Articles

Restart of natural circulation in a PWR–PKL test results and s-relap5 calculations

At reduced primary water inventory in a pressurized water reactor (PWR) with heat transfer to the secondary by means of reflux-condensation (i.e. after a small-break loss-of-coolant-accident), condense with low boron concentration may accumulate in the primary system. The restart of natural circulation after the refilling of the primary can then lead to displacement of low borated water into the reactor pressure vessel and finally into the core, which can result in a recriticality of the reactor. The restart of natural circulation was investigated in the test facility PKL, which is operated by Framatome ANP GmbH (former Siemens/KWU) in Erlangen/Germany. The test results as well as the results of a post-test calculation using the computer code s-relap5, which are both subjects of this presentation, demonstrated that despite symmetrical injection of emergency core coolant (ECC), the restart of natural circulation takes place at different times with varying intensities in the individual loops. In a PWR, this would favor the mixing of water flows with different boron concentrations in the reactor pressure vessel downcomer, which again would reduce the risk of a recriticality transient.

A simplified model of the BWR depressurization transient

A simplified model is proposed to evaluate the BWR plant depressurization transient subsequent to the activation of a single safety relief valve. The model is validated with a series of RELAP5 Mod 3 calculations yielding reasonable comparisons. The depressurization behavior of the 26 US BWR plants is analyzed with the model and the relevant parameters controlling the process are identified. Of these, four are plant-specific and their effect on the final outcome of the depressurization transient is quantified. The minimum vessel inventory at the activation of the low pressure emergency core coolant injection is selected as the figure of merit. Core power is the only parameter that significantly affects the outcome of the transient. The role of the low pressure injection and of the discharge port cross-sectional area is discussed in detail. The effect of operator actions on the control rod drive (CRD) injection flow is also discussed.

Condensation modelling for ECC injection

During a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR) important condensation phenomena take place in the neighborhood of the Emergency Core Cooling (ECC) injection ports. Direct contact condensation has been a central issue for thermalhydraulics codes such as CATHARE (Code for Analysis of ThermalHydraulic during an Accident and for Reactor safety Evaluation). The COSI experimental program was developed to simulate and study ECC injection behaviour. ECC injection has a strong local effect which requires a specific modelling. The condensation rate is controlled by the turbulent heat transfer in the liquid. A model directed toward a physical based approach is developed which incorporates the effects of such parameters as jet diameter, conduit size and flow distribution. The predicted and experimental condensation rates are compared for different test conditions. An experimental test series was devoted to study the effects of non-condensable gases on the condensation rate. An analysis of the results and a comparison with the predicted values is presented. The objective of the present study is to describe the physical mechanisms involved at the ECC injection, but also to point out some experimental and numerical limitations when modelling complex physical phenomena such as direct contact condensation.

Assessment of TRAC-PF1 Condensation Heat Transfer Model for Analysis of ECC Water Injection Transients

Direct contact condensation of vapor occurs after injection of emergency core cooling (ECC) water during a loss-of-coolant accident (LOCA) in a light water reactor (LWR). Despite that vapor condensation may have large influences on system thermal-hydraulic responses, it is difficult to predict the condensation rate accurately during ECC water injection. This difficulty sometimes results in considerable discrepancy between predicted and measured system overall transients. For instance, analyses of PWR reflood tests with the TRAC-PF1/MOD1 code predicted non-physical large flow oscillations because the code largely overpredicted the condensation rate. In an attempt to resolve this shortcoming of TRAC, a sensitivity study was carried out on the condensation heat transfer Stanton number, that is assumed to be a constant in TRAC, by analyzing the broken-loop hot leg responses to ECC injection in a combined injection test conducted at the Cylindrical Core Test Facility (CCTF). Based on this sensitivity study, it is recommended to use a Stanton number of 0.0035, which is about six times smaller than the value used in the current version of TRAC, to predict LWR system responses during an ECC water injection transient.

Assessment of TRAC-PF1 condensation heat transfer model for analysis of ECC water injection transients.

Direct contact condensation of vapor occurs after injection of emergency core cooling (ECC) water during a loss-of-coolant accident (LOCA) in a light water reactor (LWR). Despite that vapor condensation may have large influences on system thermal-hydraulic responses, it is difficult to predict the condensation rate accurately during ECC water injection. This difficulty sometimes results in considerable discrepancy between predicted and measured system overall transients. For instance, analyses of PWR reflood tests with the TRAC-PF1/MOD1 code predicted non-physical large flow oscillations because the code largely overpredicted the condensation rate. In an attempt to resolve this shortcoming of TRAC, a sensitivity study was carried out on the condensation heat transfer Stanton number, that is assumed to be a constant in TRAC, by analyzing the broken-loop hot leg responses to ECC injection in a combined injection test conducted at the Cylindrical Core Test Facility (CCTF). Based on this sensitivity study, it is recommended to use a Stanton number of 0.0035, which is about six times smaller than the value used in the current version of TRAC, to predict LWR system responses during an ECC water injection transient.

BWR Loss-of-Coolant Accident Tests at ROSA-III with High Temperature Emergency Core Coolant Injection

The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-III integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests.

BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection.

BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection.

LOFT experiment LP-02-6 analysis by RELAP5/MOD2 code with improved minimum film boiling temperature.

Groeneveld-Stewart's minimum film boiling temperature correlation was incorporated into the RELAP5/MOD2 code in order to explicitly define the minimum film boiling temperature. The transition boiling curve in the code was also modified. The Loss-of-Fluid Test (LOFT) experiment, Experiment LP-02-06 which was a cold-leg double-ended break LOCA experiment with minimum emergency core coolant injection, was analyzed with the modified RELAP5/MOD2 code. The modified RELAP5/MOD2 code well calculated system transients including the rod surface temperature transient. The temporary rewetting of rods in the early phase of blowdown, which had not been predicted by the original RELAP5/MOD2 and other codes, was predicted by the modified RELAP5/MOD2 code.

Analyses of ROSA-III Break Area Spectrum Experiments on a Boiling Water Reactor Loss-of-Coolant Accident

The ROSA-III program has conducted system effects tests on the thermal-hydraulic response of a boiling water reactor during a loss-of-coolant accident. The performance of the emergency core cooling systems was of particular interest. As part of this program, ten tests were conducted with (a) a simulated pipe rupture located at the recirculation pump suction line, (b) a spectrum of break area ranging from 0 to 200% of scaled pipe cross-sectional area, and (c) an unavailable high-pressure core spray (HPCS) system. In these tests the pressure vessel depressurized (a) due to the actuation of the automatic depressurization system for scaled break areas of 50%, and (c) due to both for the intermediate break areas between 5 and 50%. Vessel depressurization enabled the injection of emergency core coolant from the low-pressure core spray and low-pressure coolant injection system and thus led to safe recovery without HPCS. The behaviors of the vessel pressure, core mixture level, and core temperatures were fairly well reproduced by the THYDE-B1 code, based on simplified lumpedparameter models for the wide spectrum of break areas investigated.

An Analytical Study of a Small-Break Loss-of-Coolant Accident with Upper Head Injection

Upper head injection (UHI) is an emergency core coolant (ECC) system design that injects subcooled water into the upper head of the reactor vessel in a pressurized water reactor. An analysis has been performed that investigates the effects of UHI on small-break transient behavior. The analysis consists of several RELAP5/MOD1 computer code calculations, which have been compared to experimental data from a series of small-break loss-of-coolant accident simulations, performed in the Semiscale Mod-2A system. Small-break transient phenomena were calculated not to be significantly affected by the introduction of subcooled liquid into the vessel upper head. Nonequilibrium effects were minimal and limited to the period of ECC injection. The analysis covered a range of small-break sizes, and the severity of the transients (in terms of minimum core coolant level) was calculated to be a maximum (with or without UHI) for a cold leg break size of about 5.0% of the cold flow area. For all break sizes, UHI was calculated to increase the margin against core uncovery. The calculated hydraulic phenomena and specific fluid conditions were generally in good agreement with data. The calculated relative magnitudes of important phenomena were preserved over the break size spectrum.

A Model of Countercurrent Steam-Water Flow in Large Horizontal Pipes

AbstractA numerical study was performed to derive a model of countercurrent steam-water flow in large horizontal pipes, with application to the emergency core cooling (ECC) system of a pressurized water reactor. The purpose of the study was to provide data from which simple correlations could be obtained to describe mass, momentum, and energy exchange between the phases during hot leg ECC injection. It was assumed that steam, driven by a pressure drop from the upper plenum to the ECC injection port, flows counter to the subcooled ECC water. Several series of calculations were performed to determine the sensitivity of the ECC flow velocity at the entrance to the reactor vessel to the pressure drop, for several values of the mass and momentum exchange coefficients used in the numerical method. The results were consistent with those obtained from solution of the mixture equation, which did not involve interfacial drag or condensation.

Upper plenum dump during reflood in PWR loss-of-coolant accident.

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized e...

TRAC Analysis of Loss-of-Fluid Test Non-Nuclear Test L1-4

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory as an advanced best-estimate computer program for analyses of postulated accidents in light water reactors. As a part of the TRAC verification effort, a posttest analysis of loss-of-fluid test (LOFT) non-nuclear test L1-4 has been conducted. The results of this analysis show that TRAC accurately predicts the thermal-hydraulic response of the entire LOFT system, including the delayed effects from emergency core coolant injection into the cold leg.

Steam-Water Mixing in Nuclear Reactor Safety Loss-of-Coolant Experiments

Computer models used to predict the response of reactors to hypothetical accidents necessarily incorporate approximating assumptions. When attempting to verify the models by comparing predicted and measured responses in test facilities, there must be confirmation that these assumptions are realistic. Recent experiments in US facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. A noninstantaneous condensation model is described that will be incorporated into later versions of the RELAP4 code, allowing more realistic preditions of decompression rate during the emergency core coolant injection phases of a loss-of-coolant accident.

The Sensitivity of Emergency Core Coolant Bypass and Lower Plenum Refill to Apparatus Scale Size and Lower Plenum Pressure in a Pressurized Water Reactor

A series of numerical calculations was performed to study the effect of apparatus scale size on the magnitude and duration of emergency core coolant (ECC) bypass and the time delay for refill of the lower plenum of a pressurized water reactor during a hypothetical loss-of-coolant accident. These results indicate that for certain idealized flow and thermal conditions, flow similarity can be obtained at all scale sizes, but that for more realistic conditions, the effects of apparatus scale size and lower plenum pressure on ECC bypass and lower plenum refill can be large. In particular, the duration of ECC bypass and the time delay for refill appear to be more sensitive to momentum exchange at full scale and high lower plenum pressure than they are at 2/15 scale and low pressure. The sensitivity to mass exchange, ECC subcooling, and wall heat transfer decreases with increasing scale and lower plenum pressure. The effect of introducing steam, rather than air, into the downcomer through the broken ECC injection port when the pressure in the downcomer falls below that in the containment vessel is to decrease the rate of lower plenum refill.

Experimental Studies of The Effect of Flow Restrictions in a Small Rod Bundle Under Emergency Core Coolant Injection Conditions

Experimental Studies of The Effect of Flow Restrictions in a Small Rod Bundle Under Emergency Core Coolant Injection Conditions