Emergency Coolant Injection Research Articles

Hydrogen distribution in a BWŔs RPV and Mark II drywell during the progression of a severe accident with and without emergency coolant injection

This study is directed to determine potential risks associated to hydrogen concentrations in an originally fully inertized drywell of a BWR Mark II primary containment, as a consequence of the progression of a severe accident. An unmitigated case and a case with partial recovery of emergency coolant injection are studied. The base case accident scenario to determine hydrogen in-core generation and release rates to drywell is the Large Break Loss-of-Coolant Accident of a BWR, occurring simultaneously with a total loss of power. Two alternative scenarios are simulated by allowing the recovery of the High Pressure Coolant System during the progression of the base case scenario. The code RELAP/SCDAPSIM is used to simulate the different stages in the accident scenarios, and the code GASFLOW was used to compute the hydrogen distribution in a fully pre-inertized drywell. Results of hydrogen and fission product generation in core, as well as state of core degradation are presented. Then, a comparison of results of hydrogen concentration in drywell between the base case and the cases with coolant injection are presented. Results show that, for the base case, by the time of reaching primary containment venting pressure setpoint, the drywell upper zone reaches about 9 % of hydrogen concentration. For the other two cases, the resulting peak hydrogen concentrations were 12 % and 15 %.

BEPU analysis of a CANDU LBLOCA RD-14M experiment using RELAP/SCDAPSIM

A key element of the safety analysis is Loss of Coolant Analysis (LOCA) which must be performed using system thermal-hydraulic codes. These codes are extensively validated against separate effect and integral experiments.RELAP/SCDAPSIM is one such code that may be used to predict LBLOCA response in a CANDU reactor. The RD-14M experiment selected for the Best Estimate Plus Uncertainty study is a 44 mm (22.7%) inlet header break test with no Emergency Coolant Injection. This work has two objectives first is to simulate pipe break with RELAP and compare these results to those available from experiment and from comparable TRACE calculations. The second objective is to quantify uncertainty in the fuel element sheath (FES) temperature arising from model coefficient as well as input parameter uncertainties using Integrated Uncertainty Analysis package.RELAP calculated results are found to be in good agreement with those of TRACE and with those of experiments. The base case maximum FES temperature is 335.5 °C while that of 95% confidence 95th percentile is 407.41 °C for the first order Wilk's formula. The experimental measurements fall within the predicted band and the trends and sensitivities are similar to those reported for the TRACE code.

Experimental and Numerical Investigation into Temperature Distribution of a Simulated PHWR Coolant Channel Under Heatup Condition

The postulated dual-failure accident, i.e., loss of primary coolant flow along with impairment of the emergency coolant injection system, leads to peak fuel temperatures. It is well known that the temperature of the fuel assemblies is one of the significant factors that affect the outcome of an accident. Therefore, the present work aims to thoroughly investigate the thermal response of a single channel under postulated accident conditions. An experimental system was developed to capture the steady-state heat and temperature distribution in a representative 37-element fuel channel for a decay heat of 6.13 kW. Ohmic heating of the fuel rod simulators (FRSs) mimicked the generation of radioactive decay heat. Numerical simulation was also performed using the Fluent 19.1® code, and the discrete ordinates method was used to solve the radiative transfer equation. Based on the experimental results and the simulation results, it was found that the maximum Zircaloy-4 cladding temperature ≈850°C to 870°C was in the center ring. The temperature was found to vary around the circumference for each of the FRSs. Furthermore, the outer ring FRSs that had the lowest temperature developed the highest circumferential temperature gradient. In the pressure tube, the average circumferential temperature gradient obtained from the experiment and the simulation was 3.76°C/radian and 3.85°C/radian, respectively. Between the calandria tube and the moderator, the heat transfer coefficient was estimated to be around 822.3 W/m2‧K.

Validation of computer code ‘ATMIKA’ against RD-14M Small Break LOCA experiments

The simulation of Small Break Loss-of-Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility is performed under the auspices of International Atomic Energy Agency (IAEA) as an International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the in-house developed system thermal–hydraulic neutronic computer code ‘ATMIKA’, extensively used to analyze postulated events in Indian PHWRs. RD-14M is an 11MW, full-elevation-scaled extensively instrumented thermal hydraulic Canadian test facility, possessing most of the key components of a CANDU (CANada Deuterium Uranium) primary heat transport system (PHTS). The loop configuration is similar to figure-of-eight geometry of a typical CANDU circuit and it is intended to reproduce the important geometric features of a reactor PHTS and the appropriate operating conditions. ‘ATMIKA’ prediction and its comparison against SBLOCA experimental results are compared in this paper. A specific SBLOCA experiment ‘B9006’ is selected for the Computer code ‘ATMIKA’ predictions. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection (ECI) and represents most complete SBLOCA test conducted in RD-14M that includes all the phases of the transient (blow-down, high-pressure ECI, secondary pressure ramp (crash cool), refill, low pressure ECI, exponential pump ramp, and natural circulation). This simulation demonstrates that ‘ATMIKA’ is adequately capable of predicting the break discharge, PHTS depressurization, channel flow rate, channel voiding, fuel sheath temperatures and high pressure core injection flow through ECCS accumulator and initiation of low pressure ECI for test B9006. ‘ATMIKA’ predicted results are compared with experimental results and it is seen that predicted results for all phases of transient are in good agreement with experimental results.

원자력발전소의 비상냉각수 주입을 위한 초고압 펌프시스템 개발에 관한 연구

원자력발전소의 비상냉각수 주입을 위한 초고압 펌프시스템 개발에 관한 연구

Phenomena identification and ranking table for thermal-hydraulic phenomena during a small-break LOCA with loss of high pressure injection

Currently Appendix K of 10 CFR 50 is used in the United States to evaluate models for the emergency cooling systems of light-water reactors. To assure that these models are accurate enough to ensure that the cooling systems are satisfactory, code scalability, applicability and uncertainty methodologies are used to evaluate the uncertainty in system analysis code predictions due to the various models. One cornerstone of this methodology is the development of the Phenomenon Identification and Ranking Table (PIRT), which summarizes the thermal-hydraulic phenomenon associate with a particular accident scenario and ranks their importance in determining the effectiveness of core cooling and the parts of the accident for which the phenomenon may be important. In this paper the PIRT developed by the Institute for Nuclear Safety Systems for a small-break LOCA with loss of high-pressure emergency coolant injection is analyzed in detail and several modifications are proposed based on a mechanistic understanding of the phenomenon involved. The resulting PIRT should provide a more accurate guide for model evaluation and development in advanced thermal-hydraulic system analysis codes.

An open calculation of RD-14M small-break LOCA experiments using CATHENA code

The open calculations of two Small Break Loss of Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility are performed as an International Atomic Energy Agency (IAEA) activity of International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the thermal–hydraulic analysis code against qualified data for Heavy Water Reactor (HWR) systems. This ICSP started with the first meeting of the participants in Vienna in 2007 November. Two tests were selected for this activity, test B9006, a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection, performed in 1990, and test B9802, a 3-mm inlet header break experiment, performed in 1998, to provide data with full channel power to study boiling in the channels and condensation in the steam generators. The previously performed blind simulations demonstrated that the CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. However, several significant discrepancies between the code predictions and the measurement data were noted, and attributed to the input errors and the code models relevant to post-dryout (PDO) heat transfer phenomena in the heated channels of test B9802. Therefore, the open calculations are performed to allow correction of previous blind calculations in the present study. These open calculations include the Senaratne and Leung model with the optimized coefficients of its correlation for the test B9802, which can predict more accurately the onset of sustained dryout on the surface of the heater rods. The improvement of the open calculation results are shown by comparing them with the test data and also with the blind calculation results.

Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions

One of the critical issues for reactor pressure vessel (RPV) structural integrity is related to the pressurized thermal shock (PTS) event. Therefore, within the framework of safety assessments special emphasis is given to the effect of PTS-loadings caused by the nonuniform azimuthal temperature distribution due to cold water plumes or stripes during emergency coolant injection. This paper describes the method used to predict the thermal mechanic boundary conditions (system pressure and wall temperature). Using a system code the pressure and global temperature distributions were calculated, systematically varying the leak size and the location of the coolant water injection. Spatial and temporal temperature distributions in the main circulation pipes and at the RPV wall were predicted by mixing analyses with a computational fluid dynamics (CFD) code. The model used for these calculations was validated by post-test calculations of a UPTF (upper plenum test facility) experiment simulating cold leg injection during a small break loss of coolant accident (LOCA). Comparison with measured temperatures showed that the modeling used is suitable to obtain enveloping results. Fracture mechanics analyses were carried out for circumferential flaw sizes in the weld joint near the core region and between the RPV shell and the flange, as well as for axial flaws in the nozzle corner. Stress intensity factors KI were calculated numerically using the finite element program ansys and analytically on the basis of weight and polynomial influence functions using stresses obtained from elastic finite element analyses. Benchmark tests revealed good agreement between the results from numerical and analytical calculations. For all regions of the RPV investigated and the most severe transients it was demonstrated that a large safety margin against brittle crack initiation exists and brittle fracture of the RPV can be excluded.

A containment analysis for SBLOCA in the refurbished Wolsong-1 Nuclear Power Plant

A small break leading to a loss of coolant accident (SBLOCA), being one of the topic accidents in the nuclear plant diagnosis in recent years, has been analysed and evaluated for the refurbished Wolsong-1 Nuclear Power Plant (NPP). The Industry Standard Toolset (IST) codes developed by CANDU (Canadian Deuterium Uranium reactor) Owners Group and updated models including design change parameters are applied newly to the event analyses. The GOTHIC code has been used for the containment thermal-hydraulic analysis of Wolsong-1. Also, the SMART-IST code fitted in the Iodine Chemistry (IMOD-2) model has been used to predict nuclide behavior within the containment considering various aspects. The IMOD-2 was incorporated into SMART-IST as a module dealing with chemical transformations and mass transfer of iodine species in containment. IMOD-2 model is very sensitive to paint and chemicals. The parametric studies for the IMOD-2 model are performed to decide the analysis value set. The iodine release amount increases as the paint thickness increases. But, the iodine release amount increases as the water pH (dousing and primary heat transport (PHT)) decreases. The developed containment analysis methodology and the results of SBLOCA without Emergency Coolant Injection (ECI) are presented herein. Under the most heat-up conditions, the radionuclide release from the failed fuel into the containment and subsequently to the environment is such that the radioactive doses to the public are below the acceptable limits.

Scoping assessment of design characteristics for an enhanced calandria tube

Over the lifetime of a CANada Deuterium Uranium (CANDU) type reactor, the pressure tubes and calandria tubes undergo creep deformation via static, dynamic and thermal stresses accelerated by neutron bombardment. Creep deformation leads to fuelling issues, potential contact between the calandria tube (CT) and the liquid injection shutdown system or between the CT and the pressure tube (PT). As such, this aging phenomenon limits the lifetime of these components. Also, in the event of Loss of Coolant Accident (LOCA) and Loss of Emergency Coolant Injection (LOECI) scenarios, PT/CT contact may occur and if sufficient cooling is not provided, PT/CT rupture may also occur. Conceptual designs were assessed to determine their potential for reducing the effects of aging by improving CT rigidity and thermal performance of the CT. Two different design options for a CT have been investigated using numerical simulation techniques. The CT design options include fins and ribs of different sizes and combinations. The fins and ribs provide improved structural integrity and improved thermal performance over the reactors lifetime. Analyzed results have shown that the design options yield an increased overall strength with a minimal impact on fuel efficiency. The analysis has determined that the finned design option is superior in terms of CT strength enhancement yet the ribbed design is superior for improving heat transfer in accident scenarios.

A blind simulation of RD-14M small-break LOCA experiments using CATHENA code

The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down. These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena.

Adapting and applying SCDAP/RELAP5 to CANDU in-vessel retention studies

Abstract In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the long-term coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris.

Counter-current flow limitations during hot leg injection in pressurized water reactors

A one-dimensional model is presented to predict counter-current flow limitations during hot leg injection in pressurized water reactors. Different from previous models, it may also be applied in case of high Froude numbers of the liquid flow, such as to be expected in the case of emergency coolant injection through the hot leg. The model has been verified with an extensive experimental program performed in the WENKA test facility at the Forschungszentrum Karlsruhe. Typical flow regimes were investigated for a wide range of flow conditions, simulated with air and water at ambient pressure and temperature, in a simplified Pressurized Water Reactor (PWR) hot leg geometry. Depending on the water and air flow rates, flow phenomena such as a hydraulic jump and flow reversal were experimentally observed. The theoretical model shows that not only the nondimensional superficial velocities of liquid and gas, but also the Froude number of the liquid at the injection point and the Reynolds number of the gas play an important role for the prediction of flow reversal. In case of a high liquid inlet Froude number, a flow reversal could only be observed if the liquid flow became locally subcritical, i.e. if a hydraulic jump occurred in the channel. The flow reversal is predicted by the presented model with good accuracy.

Modelling of thermal and flow stratification for reactor pressure vessel pressurised thermal shock

A Pressurised Thermal Shock (PTS) occurs when the hot vessel wall of a pressurised water reactor is exposed to low-temperature, high-pressure fluid during a transient such as Emergency Coolant Injection (ECI) following a Loss of Coolant Accident (LOCA). Overcooling events have the potential to threat safety because the resulting high thermal stress, along with irradiation induced loss of ductility and stress-concentrating faults, or macro-cracks, can result in a through-the-wall crack propagation and reactor pressure vessel failure. The problem of PTS therefore has been studied extensively. This paper presents the development and application of an analytical thermal hydraulic model for predicting the growth of PTS in the downcomer in case of ECI. A preliminary assessment of pressure vessel integrity is also presented and discussed.

RELAP5 Simulation of Thermal-Hydraulic Behavior in a CANDU Reactor—Assessments of RD-14 Experiments

The critical reactor header break and the thermosiphoning experiments in the RD-14 test facility were simulated with the RELAP5/MOD3.1 code. The RELAP5 code has been developed for best-estimate transient simulation of pressurized water reactors and associated systems, but it has not been assessed for a Canada deuterium uranium (CANDU) reactor. Therefore, this study has been initiated with an aim to identify the code applicability in a CANDU reactor by simulating some of the tests performed in the RD-14 facility. The RD-14 test facility at Whiteshell Nuclear Research Establishment is a full-scale pressurized-water loop. The RD-14 is not a scale model of any particular CANDU reactor. Rather, it possesses many geometric features of a CANDU reactor heat transport system and is capable of operating at conditions similar to those expected to occur in a reactor under normal operation and some postulated accident conditions. In this study, two critical reactor header break tests (B8711 and B8713) and three thermosiphoning tests (T8513, T8515, and T8517) were analyzed with the RELAP5 code. The results were compared with experimental data and those of CATHENA performed by Atomic Energy of Canada Ltd. The RELAP5 analyses demonstrate the code’s capability to predict reasonably the main phenomena occurring in the transient, in both the qualitative and the quantitative view. However, some discrepancies after the emergency coolant injection for the critical break case and also related to the behaviors of the mass flow rate and the primary pressure for the thermosiphoning case were observed.

Contact Conductance between Cladding/Pressure Tube and Pressure Tube/Calandria Tube of Advanced Thermal Reactor (ATR)

In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure tube, and the same between pressure and calandria tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure tube made of Zr-2.5%Nb and calandria tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental results without crud compare favorably with Tachibana's correlation. The experimental contact conductances with crud are found to be lowered. Since Tachibana's correlation was not applicable to the present situation, the empirical correlation was proposed. The correlation fitted well with the experimental values.

Contact Conductance between Cladding/Pressure Tube and Pressure Tube/Calandria Tube of Advanced Thermal Reactor (ATR).

In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure tube, and the same between pressure and calandria tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure tube made of Zr-2.5%Nb and calandria tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental results without crud compare favorably with Tachibana's correlation. The experimental contact conductances with crud are found to be lowered. Since Tachibana's correlation was not applicable to the present situation, the empirical correlation was proposed. The correlation fitted well with the experimental values.

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR.

LOCA assessment experiments in a full-elevation, CANDU-typical test facility

The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined.

Degraded Cooling in a CANDU Reactor

Loss of coolant accident (LOCA) analysis for a Canada Deuterium Uranium (CANDU) reactor considers a wide range of postulated break sizes and locations in the heat transport piping. Coincident failure of the emergency coolant injection system to operate on demand must also be considered. The unique features of the CANDU core and heat transport system, and how these features affect the response of the system to a LOCA, are described. The possible range of behavior of the fuel and fuel channels following a LOCA is discussed in terms of the maximum fuel temperatures that could occur and also in terms of the potential for breaching the core pressure boundary (in the case of CANDU, this boundary comprises a large number of horizontal pressure tubes, each containing horizontal fuel bundles). It is concluded that fuel temperatures remain well below the UO/sub 2/ melting temperatures and that the integrity of the pressure tubes is maintained for all postulated LOCAs.