RELAP5/MOD3.2 investigation of a VVER-1000 MCP switching on Benchmark problem

This paper provides comparisons between experimental data of Kozloduy NPP “MCP switching on when the other three MCP are in operation”, with Relap5 calculations. The investigated thermal-hydraulic driven transient is characterized by spatially dependant non-symmetric processes. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation. The event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which leads to insertion of positive reactivity due to the modeled feedback mechanisms. The main purpose of this investigation was to improve the discrepancy between the calculations and the plant data. The sensitivity calculation investigates the mixing in reactor vessel and influence of heat structure on the hot legs temperature. The areas of improvements to the Relap5 model are:

• •The non-symmetrical mixing in downcomer and reactor vessel annular exit.
• •The influence of heat structure temperature on the time delay for equipments measurements.
• •Investigation of pressurizer water level – using the hot legs temperature correction.

The RELAP5/MOD3.2 model of Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios have been developed and validated in the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS) Sofia, and Kozloduy NPP. The model provides a significant analytical capability for the specialists working in the field of NPP safety.This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of the VVER-1000 against the experimental transient data received from the Kozloduy NPP Unit 6. The comparisons indicate good agreement between the RELAP5 results and the experimental data. The sensitivity investigation improves the discrepancy between the calculation and the plant data.This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Validation of RELAP5/MOD3.2 model on trip off one main coolant pump for VVER 440/V230

This paper presents the validation of RELAP5/MOD3.2 model of the VVER 440 for Nuclear Power Plant (NPP) in the analysis of the following transient: “Trip off one MCP”.This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with measurement transient data received from Kozloduy NPP Unit #4. The baseline input deck for VVER440 was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events, and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP.The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparison between the RELAP5 calculations and the test data indicates a good agreement.This validation was possible through the participation of leading specialists from Kozloduy NPP and with the support of the Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Flow analyses using RELAP5/MOD3 code for OPR1000 under the external reactor vessel cooling

External reactor vessel cooling (ERVC) is considered as one of the most promising severe accident management strategies for an in-vessel corium retention (IVR). Heat removal capacity and water availability at the vessel outer surface can be key factors determining the success of ERVC measures. In this study, for the investigations on the effect of water availability in case of ERVC, flow analyses using the RELAP5/MOD3 code were performed. The analyses were focused to examine the flow behavior inside the reactor pressure vessel (RPV) insulator of the OPR1000 (Optimized Power Reactor 1000 MWe) under a cavity flooding. The current flow analyses results show that for the accident scenarios of station black out (SBO) and 9.6 in. large break loss of coolant accident (LBLOCA) under the ERVC, steam could not ventilate through the insulator and the pressure inside the RPV insulator increased abruptly. This induced a water sweep out and steam domination in the flow path inside the insulator. These flow analyses results indicate that sufficient water ingression and steam venting through the insulator can be a key factor determining the success of the ERVC in the operating nuclear power plant, OPR1000. According to the results of the sensitivity studies for the venting area, in terms of an effective flow circulation inside the insulator, an optimal venting is to assign four holes having a diameter of 0.3 m at the upper exit (hot leg level) of the insulator. And the effect of the inlet flow area at the insulator bottom is rather minor when compared to that of the outlet flow area of a steam venting.     

An assessment of a VVER-1000 core during Turbo-Generator load reduction test using RELAP5/MOD3.2 and WIMSD-5B/PARCSv2.7

Reactor dynamic tests, which are categorized as one of the power start-up test groups, are the most complex tests during commissioning of the new nuclear power plants. This paper presents the results of Turbo-Generator load reduction test as one of the reactor dynamic tests for VVER-1000/V446 unit at Bushehr Nuclear Power Plant (BNPP). In this test modeling because of the need for control rod bank worth and core reactivity coefficients, the core geometry has been modeled first by using WIMSD-5B/PARCSv2.7 codes for neutronic calculations. For performing the thermal-hydraulic analysis, the RELAP5/MOD3.2 computer code has been used. The control rod bank worth and core reactivity coefficients obtained from WIMSD-5B/PARCSv2.7 are compared with BNPP FSAR that confirm the ability and reliability of the method. Also comparison of the thermal-hydraulic core parameters obtained from RELAP5/MOD3.2 against actual plant data, indicate that this code can properly predict behavior of VVER-1000 reactor for this dynamic start-up test.     

Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code

Incorporation of full three-dimensional models of the reactor core into system thermal–hydraulic transient codes allows better estimation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations to verify and validate the capability of the so-called coupled codes technique. For these purposes appropriate Light Water Reactor (LWR) transient benchmarks based upon programmed transients performed in Nuclear Power Plants (NPP) were recently developed on a higher ‘best-estimate’ level. The reference problem considered in the current framework is a Main Coolant Pump (MCP) switching-on transient in a VVER1000 NPP. This event is characterized by a positive reactivity addition as consequence of the increase of the core flow. In the current study the coupled RELAP5/PARCS code is used to reproduce the considered test. Results of calculation were assessed against experimental data and also through the code-to-code comparison.     

基于修改后RELAP5/MOD3.2的氟盐高温堆实验回路分析

大型热工流体整体效应系统实验(CIET)台架是为模拟氟盐冷却高温堆(FHR)热工水力响应而设计的实验回路,采用DOWTHERM A模拟氟盐作为冷却剂。通过在RELAP5/MOD3.2程序中加入DOWTHERM A物性参数以及传热关系式,计算FHR实验回路CIET在两种工况下的热工水力行为,并与实验结果进行对比,计算工况包括强迫循环条件与自然循环条件。计算结果表明:在强迫循环条件下,堆芯热量主要靠盘管式空气换热器(CTAH)排出,堆芯进出口冷却剂温度及CTAH出口冷却剂温度与实验值符合良好,CTAH进口冷却剂温度与实验值有些微偏差;在自然循环工况中,堆芯热量主要通过DHX与堆芯辅助冷却系统(DRACS)回路的换热带走,DHX及DRACS的流量与实验值接近,相对误差在10%左右,验证了修正后RELAP5/MOD3.2的正确性。     

Direct-contact condensation heat transfer model in RELAP5/MOD3.2 with/without noncondensable gases for horizontally stratified flow

Direct-contact condensation experiments on atmospheric steam and steam/air mixture on subcooled water flowing co-currently in nearly-horizontal channels are carried out and the local heat transfer coefficients are obtained. Inlet air mass fraction in the mixture is varied by up to 50%. Based on previous and present experimental data, a direct-contact condensation database is constructed. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both co-current and counter-current experimental data. The correlation proposed by Kim predicts the database relatively well compared with that of RELAP5/MOD3.2. In the presence of noncondensable gases, RELAP5/MOD3.2 overpredicts the database with percentage errors of 55.1 and 61.1% for pipe inclination angles of θ=2.1° and θ=5.0°, respectively. The UCB correction factor is modified to consider the effects of noncondensable gases on heat transfer coefficients. When Kim's correlation is substituted with the Dittus–Boelter type correlation in RELAP5/MOD3.2 and a modified correction factor is used, the prediction errors are greatly reduced to 20.7 and 28.8% for inclination angles of θ=2.1° and θ=5.0°, respectively.     

RELAP5/MOD3.2竖直管束外大容积沸腾换热模型适用性分析

以浸没在大容积水箱内的非能动余热排出换热器为研究对象,采用实验数据对RELAP5/MOD3.2程序计算竖直管束外大容积沸腾换热适用性进行校核。在充分考虑大容积水箱内流体沿管束轴向的自然对流以及径向的气液间热量交换基础上,合理建立了沸腾换热回路节点划分模型。将计算结果与实验数据进行对比,发现沸腾换热系数的计算值与实验值最大相对偏差在50%以上,且沸腾换热系数随热流密度变化的趋势明显不同。由此判断,Chen关系式并不适合计算竖直管束外大容积沸腾的情况。通过与已有的大容积沸腾换热计算关系式对比,发现Kutateladze “new”公式或Rohsenow公式计算值与实验值符合较好。     

Analysis of maximum pressure in VVER1000/V446 reactor containment for LOCA and MSLB

The highest thermal-hydraulic pressure in the containment occurs when reactor coolant in the first loop and steam in the secondary loop discharge simultaneously,and when the maximum amount of energy from reactor unit enters to containment volume.In this paper,we investigate temperature and pressure variations in the VVER1000 containment compartments owing to concurrent break in the pipelines of the primary and secondary loops.A two-phase,multicellular model is applied in the presence of non-condensable gases.Convection and conduction through the main heat structures inside the containment are also considered.The predicted results agree well with available data.Maximum values of pressure and temperature in the containment are then calculated and compared to the design values.If LOCA and MSLB occur simultaneously,the maximum pressure would exceed the design value and integrity of the containment would be threatened.     

Thermal hydraulic analysis of Syrian MNSR research reactor using RELAP5/Mod3.2 code

This paper presents the development and validation of a MNSR-RELAP5 model. MNSR is a 30 kW, light-water moderated and cooled, beryllium-reflected, tank in pool type research reactor. A RELAP5 model was set up to simulate the entire MNSR system. The model represents all reactor components of primary and secondary loops with the corresponding neutronic and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents are simulated.     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).     

Evaluation of analytically scaled models of a pressurized water reactor using the RELAP5/MOD3 computer code

This paper performs analytical evaluations for the potential distortions caused by the scaled models using RELAP5/MOD3 computer codes. By use of scaling analysis, two scaled models with the same volumetric ratio are constructed for the Korean next generation reactor (KNGR), which is an advanced light water reactor. The scaling methodology adopted in this paper preserves the two-phase natural circulation similarities between prototype and scaled models. One scaled model is at full height with reduced flow area. The other model is at reduced height with reduced flow area. By using appropriate scale factors the RELAP5/MOD3 input models are developed. Then, the transient responses of the two ideal scaled models are simulated for small break loss of coolant accidents (SBLOCAs) by using the RELAP5/MOD3 computer code. The transient responses of the two scaled models are compared with those of the prototype. The results indicate that qualitative and quantitative similarities are well preserved for both models during SBLOCA with different break sizes.     

Evaluation of the RCS depressurization strategy for the high pressure sequences by using SCDAP/RELAP5

As part of the evaluation for a severe accident management strategy, a reactor coolant system (RCS) depressurization in optimized power reactor (OPR)1000 has been evaluated by using the SCDAP/RELAP5 computer code. An indirect RCS depressurization by a secondary depressurization by using a feed and bleed operation has been estimated for a small break loss of coolant accident (LOCA) without a safety injection (SI). Also, a direct RCS depressurization by using the safety depressurization system (SDS) has been estimated for the total loss of feed water (LOFW). The SCDAP/RELAP5 results have shown that the secondary feed and bleed operation can depressurize the RCS, but it cannot depressurize the RCS sufficiently enough. For this reason, a greater direct RCS depressurization by using the SDS is necessary for the 1.35 in. break LOCA without SI. A proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time from 7.5 to 10.7 h. An opening of two SDS valves can depressurize the RCS sufficiently enough and the proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time of approximately 5 h for the total LOFW. An opening of one SDS valve cannot depressurize the RCS sufficiently enough.     

Integral loop test and assessment of modified RELAP5/MOD3.3 for RCS coolant inventory during a LBLOCA

This study investigates experimentally and analytically the thermal hydraulic phenomena during the transition from design basis accident (DBA) to beyond-DBA, particularly, the depletion of core coolant inventory. To investigate the overall thermal hydraulic behavior after safety injection (SI) failure during a large-break loss-of-coolant accident (LBLOCA) in a cold leg, an integral loop test was performed at the Seoul National University Integral Test Facility (SNUF), which was scaled down to 1/6.4 in length and 1/178 in area from the advanced power reactor 1400 MWe (APR1400) according to the three-level scaling method. The plant condition at 200.0 s as the base case and those at 625.0 and 1950.0 s as test cases after the initiation of LBLOCA were applied as initial conditions in the experiments, respectively. The experimental results showed that the sweepout increased the coolant flow discharged to the break depending on the steam flow rate in intact cold legs and the initial downcomer coolant level and expedited the depletion of the core coolant inventory.In the meantime, since RELAP5/MOD3.3 uses the average properties of donor volume as those of its connected junction, this scheme causes the mass and the energy flux in a junction to be calculated incorrectly if significant phase separation occurs in the donor volume such as in the downcomer during the LBLOCA. The sweepout model was developed and implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory during the LBLOCA. To assess the applicability of the modified RELAP5/MOD3.3 to the actual system, the experiments in SNUF were simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the discharge flow rate at the break larger than that of the experiment. On the other hand, the modified one calculated the discharge flow rate more similar to that of the experiment than the original one did. As a result, the modified RELAP5/MOD3.3 reduced the coolant flow discharged to the break to delay the initiation time of heater heat-up after SI failure during LBLOCA in a cold leg. This improved RELAP5/MOD3.3 will support a more realistic thermal hydraulic analysis in an integrated analysis system.     

Development and validation of a transition boiling model for RELAP5/MOD3 reflood simulation

The current version of the RELAP5/MOD3.1 code significantly underpredicts the transition boiling heat transfer during reflooding of hot fuel rods. In order to extend the code’s range of application for LOCA and degraded core analyses, a new transition boiling model has been developed, assessed and implemented. The model is based entirely on local state variables calculated by the code (wall and fluid temperatures, pressure, void fraction, mass flux and static quality) and does not rely on other history parameters, such as quench position or CHF and minimum film boiling temperatures. A number of separate-effect and bundle experiments are analyzed with the modified version of the code, and the predictions are compared with the ones obtained by the current version and with available experimental data. In all cases, the predictions of the improved model better fit the measured data. The shape of the new temperature curves is more physically and conceptually sound than the one calculated by the current version of the code.     

RELAP5/MOD3.3 analysis of coolant depletion tests after safety injection failure during a large-break loss-of-coolant accident

This study is concerned with development of a coupled calculation methodology with which to continually and consistently analyze progression of an accident from the design-basis phase via core uncovery to core melting and relocation. Experiments were performed to investigate the core coolant inventory depletion after safety injection failure during a large-break loss-of-coolant accident in a cold leg utilizing the Seoul National University Facility (SNUF). The SNUF is an integral test loop scaled down to 1/6.4 in length and 1/178 in area from the Advanced Power Reactor 1400 MWe (APR1400). The SNUF tests are simulated with the RELAP5/MOD3.3 code. The test results revealed that the core coolant inventory decreased five times faster during the sweepout in the downcomer than after termination of the sweepout. The sweepout was observed to take place on top of spillover from the downcomer region to expedite the depletion of the core coolant inventory. The calculation results of RELAP5/MOD3.3 deviated from the experimental data in terms of entrainment from the surface of core coolant, condensation and sweepout in the downcomer. Thereby, the core coolant level was computed to decrease faster than the measured from the experiment due to the overestimated spillover by the evaporation of the entrained droplets by the uncovered heaters. Notwithstanding the occasional disparities, the code prediction is in reasonable agreement with the overall behavior of the tests.     

Simulation of SBLOCA based on an Improved Choked Flow Model for RELAP5/MOD3 Code

This paper is a continuation of the present author’s previous publication dealing with a new choked flow model for two-phase flow. The model based on a hyperbolic one-dimensional two-fluid model, where in the momentum equations the terms representing the interfacial pressure difference has been included in lieu of the virtual mass force terms. The new choked flow model is an improvement upon the choked flow model of the current RELAP5/MOD3 code, which itself is based on the Trapp–Ransom method. The author compares the predictions of this improved model with Trapp–Ransom model and Henry–Fauske model, for an assumed flow in a vertical pipe. The author simulates a typical PWR system with a hypothetical SBLOCA as well, and compares the system behaviors predicted by RELAP5/MOD3, based on the aforementioned choked flow models. He shows that the improved choked flow model leads to better predictions.     

Effect of depressurization on reactor vessel inventory in the absence of ECCS injection

A series of tests were performed to evaluate inventory depletion as a reactor vessel undergoes depressurization in the absence of any emergency core coolant system injection (ECCS). These tests were carried out in a scaled representation of a reactor vessel which was initially filled with saturated water up to the elevation of the hot legs. Depressurization valves installed on take-off lines from the hot legs were opened and level swell ensued in the reactor vessel initiating a two-phase blowdown. This was followed by subsequent single-phase discharge transient which in some cases led to core uncovery. A combined model encompassing the two-phase and single-phase discharge portions of the transient is proposed. The inventory-versus-pressure traces obtained from the model compare well with the experimental results. These traces are discussed as bounding trajectories for a large class of small break loss of coolant accident (LOCA) transients which otherwise must be considered individually.     

Improvement of the reflood model of RELAP5/MOD3.3 based on the assessments against FLECHT-SEASET tests

The reflood model of RELAP5/MOD3.3 was assessed using eight carefully selected FLECHT-SEASET tests. Comparisons of calculated and measured peak cladding temperatures and quench times indicate that the code predicts peak cladding temperatures relatively well. However, rod quenches were predicted to occur too early.To improve the predictability for quench times, we carefully reviewed and modified the wall-to-fluid heat transfer models. The film boiling heat transfer regime was divided into three different sub-regimes, and appropriate correlations were selected for the wall-to-fluid heat transfer coefficients in each sub-regime. We modified also the wall-to-liquid radiation heat transfer model including the method for determining the droplet diameter and we corrected two minor coding errors existing in the original version. After the modifications, the same set of eight FLECHT-SEASET tests was simulated again to compare the newly predicted cladding temperatures, quench times and heat transfer coefficients with those predicted using the original version. The modifications employed in this study improve the code's predictability not only for quench times but also for peak clad temperatures. The modifications reduced the Root-Mean-Square (RMS) error in the prediction of peak clad temperatures and quench times from 48.3 to 36.1 K and 85.9 to 33.2 s, respectively.     

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.