Mid-loop Operation Research Articles

Integral-Effect Test for Loss of Residual Heat Removal System During Mid-Loop Operation in ATLAS Facility

A test called B4.2 in the OECD-ATLAS2 project was performed to simulate loss of the residual heat removal system (RHRS) during mid-loop operation (MLO) using a thermal-hydraulic (T-H) integral-effect test facility: the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). The main purpose of this test was to investigate a T-H transient in the reactor coolant system (RCS) during loss of the RHRS and to evaluate the effectiveness of reflux condensation and the capability of a safety injection tank (SIT) on shutdown coolability. The initial and boundary conditions for the B4.2 test were appropriately determined according to a state of MLO corresponding to 65 h after reactor trip in the Advanced Power Reactor 1400 MW(electric) (APR1400). During the loss of RHRS accident transient simulation, major T-H parameters such as system pressures, temperatures, and collapsed water levels in the RCS were measured, and unique T-H phenomena such as reflux/cocurrent condensations, off-take, countercurrent flow, and countercurrent flow limitation were investigated. In this paper, the overall T-H behavior in the RCS during a simulated loss of the RHRS with SITs is highlighted to provide a better understanding of T-H phenomena regarding coolability with reflux condensation.

Numerical investigation of two-component single-phase natural convection and thermal stratification phenomena in a rod bundle with axial heat flux profile

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

Analysis of severe accident progression and Cs behavior for SBO event during mid-loop operation of OPR1000 using MELCOR

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

Effects of fluid properties on interfacial and wall friction factors under counter-current flow limitation in a vertical pipe with sharp-edged lower end

A gas–liquid counter current flow may possibly occur in the U-tubes of the steam generator in a PWR due to loss of the residual heat removal during mid-loop operation. For safety analysis based on the one-dimensional multi-fluid simulation, correlations of the interfacial and wall friction factors are required. Experiments on air-glycerol water solution counter-current annular flows in a circular vertical pipe of 20 mm in diameter and 400 mm in length with a sharp-edged lower end, hence, were carried out to investigate the effect of the liquid viscosity on the flow structure and the interfacial and wall friction factors. The liquid flow rate, the pressure gradient and the liquid volume fraction were measured to evaluate the friction factors. The time-strip flow visualization technique was applied to flow images in the pipe for detailed observation. The obtained conclusions are as follows: (1) the flows under CCFL condition in the present study can be classified into two regimes, i.e. the rough film with disturbance waves forming at the lower pipe end (RF-I) and the disturbance waves flowing upward with dissipation and regeneration (RF-II), (2) the interfacial friction factor is independent of the liquid viscosity and mainly determined by the balance between the inertial and buoyant forces. The gas Froude number, JG*, is the primal dimensionless group in correlating the interfacial friction factor, (3) the wall friction factor can be correlated in terms of the liquid Froude number JL*, the liquid Reynolds number ReL and the dimensionless pipe diameter D*, and (4) the correlations of the friction factors were proposed. The applicable ranges of the correlations are 0.58 ≤ JG*1/2 ≤ 0.78, 0.01 ≤ JL*1/2 ≤ 0.10, 0.15 ≤ ReL ≤ 178 and 7.28 ≤ D* ≤ 17.79.

Validation of interfacial area concentration model for simulating bubbly and cap/slug flow behaviors in large diameter pipes using LSTF, ATLAS, and horizontal pipe experiment data

This study discusses Interfacial Area Concentration (IAC) models developed for bubbly and cap/slug flow regimes in a large diameter pipe incorporating recently developed small bubble fraction correlations in various flow conditions. The main focus of this study was the robust simulation of the behaviors of bubbly and cap/slug flows in large diameter pipes observed in Large Scale Test Facility (LSTF), Advanced Thermal-hydraulic test Loop for Accident Simulation (ATLAS), and horizontal flow test facility. Since the IAC was one of the key parameters to achieve this, an IAC model for large diameter pipes was first derived based on the existing Hibiki-Ishii IAC model for bubbly flow, and the Shen-Hibiki IAC model for cap bubbly-to-churn flow. The volume fraction of group-1 bubbles, i.e., small spherical bubbles, in the IAC model, which have been developed based on the assumption that the group-1 bubbles exponentially decay in the slug flow regime, was refined. As a result, a more physical model of the group-1 bubble number density that remains constant in the cap/slug flow regime was imposed for large diameter pipes. The validity of the IAC model for large diameter pipes in predicting the interfacial drag force between gas and liquid phases in large pipes of multiple test facilities was then examined. The validation of the proposed IAC model was conducted through simulations of 1) the loss of Residual Heat Removal (RHR) during the mid-loop operation of the LSTF, 2) the 6-inch Small Break Loss of Coolant Accident (SBLOCA) of the ATLAS, and 3) the two-phase flow behaviors under a wide range of fluid conditions in the horizontal pipe test facility. Improved results were obtained with the proposed model as it indicated that there was good agreement between the calculated results and the experimental data throughout the whole transient.

Application of BEPU method to loss of RHR event during mid-loop operation: Uncertainty quantification of RELAP5 model

In recent years, best estimate plus uncertainty (BEPU) method has become more important, and the guidelines on the process have been developed. However, there are not many specific reports on the probability distribution of the model uncertainty that greatly affects the evaluation result. Also, there are not many application cases of the BEPU method for the events of the PWR mid-loop operation which is suggested to be important in terms of probabilistic risk assessment (PRA). Therefore, this report describes the results of the quantification of the RELAP5/MOD3.2 model uncertainties related to the loss of residual heat removal (RHR) event during the PWR mid-loop operation. Based on the experimental literatures, the probability distributions of the model uncertainty were developed, which contributes to statistical analyses and research on the model reliability.

Analysis of loss of residual heat removal system during mid-loop operation of a 300 MWe PWR

Thermal-hydraulic response of a two-loop pressurized water reactor is studied on loss of residual heat removal system (RHRS) during mid-loop operation. Decay heat during mid-loop operation is removed by RHRS. On loss of RHRS, no cooling system is available to cool the core. Transient thermal-hydraulic analysis of the plant is performed using RELAP/SCDAP on loss of RHRS. For this purpose, first steady-state is achieved at mid-loop conditions. System obtains steady-state at primary coolant inventory of 58000 kg, water level of 7.8m in the reactor pressure vessel (RPV) and upper plenum pressure of 104.41 kPa. On loss of RHRS, the core uncovery time, maximum reactor system pressure and cladding temperature are determined under four accidental scenarios in absence of any mitigation measure. The results show that the core uncovery time for the open case of reactor pressure vessel top vent valve (RPV top vent valve), pressurizer relief valve (PZR valve), steam generator man way and pressurizer man way are 423.1, 214.4, 50.1, 67.2 min respectively. The effect of reflux condensation for the above accidental configurations has also been studied. It is found that reflux condensation is a prominent phenomenon to increases core uncovery time for open case of RPV top vent valve and PZR valve.

Numerical simulation of vortex in residual heat removal system during mid-loop operation

Abstract Free surface vortex, especially in a pattern of full air core vortex, causes serious security problems in the residual heat removal system. Vortex in a T-junction pipe system was numerically investigated under the premise that the simulation results are in agreement with the experimental results. First, four typical states were obtained in the formation and evolution process of vortex, namely, surface dimple, critical submergence state, air entrainment vortex state, and large air entrainment state. With increasing Froude (Fr) number, the time from an initiate flow field to the critical submergence state decreased, and the time interval between critical submergence state and large air entrainment state increased. Finally, the variation law of vortex intensity was analyzed during the evolution process. The vortex intensity initially increased with time and reached its maximum value and then decreased until it reached a minimum value before the large air entrainment state. With the increase of Fr, the maximum values of vortex intensity gradually increased.

Application of the Statistical Safety Evaluation Method to Reflux Cooling in the Loss-of-RHR Event During Mid-Loop Operation

Application of the statistical safety evaluation method to the analysis for reflux cooling in the loss of residual heat removal system (RHR) event during mid-loop operation was conducted with RELAP5/MOD3.2 as the best-estimate code. The evaluation parameters for this event are the reactor coolant system (RCS) pressure and reactor vessel (RV) water level. A phenomena identification and ranking table was developed previously to identify important phenomena that have a large influence on the evaluation parameters. The uncertainties of the analytical models which are related to the important phenomena were quantified to use as input for the statistical analysis. Actual plant sensitivity analyses were carried out to confirm the influence of the uncertainty which was quantified. Validation analysis of the RELAP5/MOD3.2 code was carried out for the reflux cooling test conducted with the ROSA-IV Large-Scale Test Facility to confirm that the experimental RCS pressure results were within the range of the uncertainty of the analytical models. A statistical analysis for 93 cases was carried out for a four-loop plant to evaluate the RV water level in the steady state and the maximum RCS pressure with 95% cumulative probability and 95% confidence. As a result, it was confirmed that reflux cooling using two cooling steam generators was successful.

ASTEC V2.0 computational evaluation of source term and its chemical forms under accidental conditions during mid-loop operation

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO 4 is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I 2 + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I 2 persistence). Creation of gaseous RuO 4 is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram.

Numerical analysis of steam–air behavior in a pressurizer during reflux cooling

During reflux cooling, proper evaluation of behavior of accumulated non-condensable gases in the steam generator (SG) U-tubes is important to predict the performance of the reflux cooling. Non-condensable gases are present in the pressurizer and the possibility of migration of air in the pressurizer to the SG U-tubes is not well known. Steam and air behavior in the pressurizer during reflux cooling was, therefore, analyzed numerically using FLUENT 6.3.26 and the possibility of migration of air to the hot leg was investigated. For the analysis, the pressurizer of the ROSA-IV/LSTF experiment was employed as a calculation domain, since experimental data about the loss of the residual heat removal event during mid-loop operation are available. Two stages were assumed. (1) Phase 1: latent heat accumulated in the wall of the pressurizer and was eventually released to the outside. (2) Phase 2: the wall was heated up to the saturated steam temperature, and only heat loss to the outside occurred. The prediction indicated that in Phase 1, the air did not migrate to the surge line in either laminar or turbulent flow calculations, while in Phase 2 the air migrated into the hot leg only in the laminar flow calculation. Judging from a previous experiment on an axisymmetric free jet, the flow pattern in the pressurizer seems to be turbulent. In addition, a comparison of the analytical results of the fluid temperatures near the wall of pressurizer with ROSA-IV/LSTF experiment results indicated that the turbulent flow calculation results were more realistic. It was therefore concluded that the turbulent flow calculation was more reasonable and the possibility of migration of air to the hot leg was low in a pressurizer during reflux cooling.

Countercurrent Gas-Liquid Flow in a PWR Hot Leg under Reflux Cooling (I) Air-Water Tests for 1/15-Scale Model of a PWR Hot Leg

In the case of loss of residual heat removal systems under mid-loop operation during shutdown of a PWR plant, reflux cooling by a steam generator (SG) is expected, and steam generated in the reactor core and water condensed in the SG form a countercurrent flow in a hot leg. The flow is highly complicated because the hot leg consists of a horizontal pipe, an elbow, and an inclined pipe. In this study, a scale model of the PWR hot leg (1/15th of the actual plant size) is used to investigate flow patterns and characteristics of countercurrent flow limitation (CCFL). The following conclusions are obtained. (1) The effects of gas volumetric flux JG and water volumetric flux JG on flow pattern in the hot leg were made clear. Flow patterns in the elbow and inclined section are strongly affected by those in the horizontal section. (2) When the flow rate of the supplied water is constant, the value of JG at the onset of transition from stratified flow to wavy flow obtained by increasing JG is larger than that at the onset of transition from wavy flow to stratified flow obtained by decreasing JG . (3) CCFL characteristics obtained by increasing JG differ from those obtained by decreasing JG . CCFL data obtained by decreasing JG agree well with available data. (4) The boundary between wavy flow and stratified flow in the horizontal section obtained by decreasing JG agree well with the CCFL characteristics.

Countercurrent Gas-Liquid Flow in a PWR Hot Leg under Reflux Cooling (II) Numerical Simulation of 1/15-Scale Air-Water Tests

When reflux cooling is executed in the case of loss of residual heat removal systems under mid-loop operation of PWR, steam generated in the reactor core and water condensed in the steam generator may form a complicated countercurrent flow in the hot leg. Numerical simulations of air-water countercurrent flows in a 1/15-scale model of the hot leg are carried out in this study to examine the capability of predicting two-phase flow patterns and CCFL characteristics in the hot leg. A three-dimensional two-fluid model implemented in FLUENT6.3.26 is used in the simulations. Good agreements between measured and predicted flow patterns and CCFL characteristics are obtained by using an appropriate set of correlations for the interfacial friction coefficient in the momentum equation of the two-fluid model.

Void Fraction in a Four by Four Rod Bundle under a Stagnant Condition

In the case of a hypothetical failure of a residual heat removal (RHR) systems under mid-loop operation, vapor generated in a reactor core forms two-phase flow in a stagnant liquid and rises the water level in the core. The vapor flows into a steam generator through a hot leg, and condenses in the steam generator. Since the flow rate of vapor from the reactor core to the hot leg depends on the water level and the void fraction α in the reactor core, the reliable analysis of the RHR failure cannot be carried out without accurately estimating the void fraction in the reactor core. Although a number of studies on void fractions in two-phase flows in rod bundles have been carried out, there are few experimental data on void fractions in rod bundles under the stagnant condition. Void fractions in four by four rod bundles under the stagnant condition were measured for a wide range of gas volume fluxes to examine the validity of available void correlations. Flow patterns were visualized by using a high-speed video camera to examine the effects of flow pattern on the void fraction. As a result, the following conclusions were obtained: (1) Dependence of the void fraction on the gas volume flux JG changed at JG ≅ 1.5 m/s due to the flow pattern transition. (2) Murase's correlation agreed well with the void fraction in the two kinds of rod bundles having different dimensions under the stagnant condition.

Countercurrent Gas-Liquid Flow in a PWR Hot Leg under Reflux Cooling (I) Air-Water Tests for 1/15-Scale Model of a PWR Hot Leg

In the case of loss of residual heat removal systems under mid-loop operation during shutdown of a PWR plant, reflux cooling by a steam generator (SG) is expected, and steam generated in the reactor core and water condensed in the SG form a countercurrent flow in a hot leg. The flow is highly complicated because the hot leg consists of a horizontal pipe, an elbow, and an inclined pipe. In this study, a scale model of the PWR hot leg (1/15th of the actual plant size) is used to investigate flow patterns and characteristics of countercurrent flow limitation (CCFL). The following conclusions are obtained. (1) The effects of gas volumetric flux JG and water volumetric flux JG on flow pattern in the hot leg were made clear. Flow patterns in the elbow and inclined section are strongly affected by those in the horizontal section. (2) When the flow rate of the supplied water is constant, the value of JG at the onset of transition from stratified flow to wavy flow obtained by increasing JG is larger than that at the onset of transition from wavy flow to stratified flow obtained by decreasing JG . (3) CCFL characteristics obtained by increasing JG differ from those obtained by decreasing JG . CCFL data obtained by decreasing JG agree well with available data. (4) The boundary between wavy flow and stratified flow in the horizontal section obtained by decreasing JG agree well with the CCFL characteristics.

Countercurrent Gas-Liquid Flow in a PWR Hot Leg under Reflux Cooling (II) Numerical Simulation of 1/15-Scale Air-Water Tests

When reflux cooling is executed in the case of loss of residual heat removal systems under mid-loop operation of PWR, steam generated in the reactor core and water condensed in the steam generator may form a complicated countercurrent flow in the hot leg. Numerical simulations of air-water countercurrent flows in a 1/15-scale model of the hot leg are carried out in this study to examine the capability of predicting two-phase flow patterns and CCFL characteristics in the hot leg. A three-dimensional two-fluid model implemented in FLUENT6.3.26 is used in the simulations. Good agreements between measured and predicted flow patterns and CCFL characteristics are obtained by using an appropriate set of correlations for the interfacial friction coefficient in the momentum equation of the two-fluid model.

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.

Simulator performance upgrade and evaluation of loss of RHR during midloop operation for Maanshan PWR plant

Because a greater risk than expected is introduced during midloop operation for typical PWRs, the performance of training simulator of Taiwan’s Maanshan 3-loop PWR plant was verified and upgraded for midloop operation (MLO) simulation. Besides, plant specific midloop abnormal operation procedure (AOP) also was quantitatively evaluated. Instead of modifying existing RCS module, a thermal-hydraulic code, namely ROSE (Reactor Outage Simulation and Evaluation) has been developed and transplanted into the training simulator of Maanshan PWR plant. A two-region approach with a modified two-fluid model was adopted as the theoretical basis of the ROSE code. The success of the simulator performance upgrade for the MLO was demonstrated by comparisons to W-GOTHIC MLO calculation as well as the original simulator performance. Moreover, regarding the evaluation of the associated AOP for MLO after loss of RHR, the most risky plant configuration as well as associated crucial timing before core uncovery was also identified by the upgraded training simulator.

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.

Analysis of a loss of residual heat removal system during mid-loop conditions at PKL facility using RELAP5/Mod3.3

When a nuclear power plant is in shutdown conditions for refuelling, the reactor coolant system water level is reduced. This situation is known as mid-loop operation, and the residual heat removal (RHR) system is used in this situation to remove the decay power heat generated in the reactor core. In mid-loop conditions, some accidental situations may occur with not a negligible contribution to the plant risk, and all involve the loss of the RHR system. Thus, to better understand the thermal–hydraulic processes following the loss of the RHR during shutdown, transients of this kind have been simulated using best-estimate codes in different integral test facilities. This paper focuses on the simulation, using the best estimate code RELAP5/Mod3.3, of the experiment E3.1 conducted at the PKL facility. This experiment consists of the loss of the RHR system when the plant is in mid-loop conditions for refuelling and with the primary circuit closed. In particular, in this experiment the physical phenomena to investigate are the mechanisms of heat removal in presence of nitrogen and the deboration in critical parts of the primary system.