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.     

A separate effect test on the cooling behavior in a pressurized water reactor core under a low reflooding rate condition by using ATLAS and RELAP5/MOD3.3 code assessment

A separate effect test was performed on the cooling behavior in a PWR core under a low reflooding rate condition by using the ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) which is a thermal–hydraulic integral effect test facility for the pressurized water reactors APR1400 and OPR1000. Although several integral tests for the reflood phase of a large break loss of coolant accident (LBLOCA) of APR1400 have been performed with the ATLAS, the previous integral effect tests for the reflood phase of a LBLOCA are not easily simulated by existing codes, such as the RELAP5/MOD3, due to a unique phenomena in ATLAS, that resulted from an injection of large amount of subcooled water onto the heated wall of which temperature was higher than the target value.     

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.     

LOCA analysis of reactor cores with tight lattice using the modified RELAP5 code

The previous paper analyzed the reflooding phase of reactor cores with tight lattice. Models calculating the wall to fluid heat transfer in the precursory cooling region and in the vicinity of the quench front were developed and validated in the previous paper (Wu et al., 2012). In this paper, these newly developed models were used to modify RELAP5/MOD3.2 in order to make the code be suitable for tight lattice. Besides, minor modifications to the wall friction model and bubbly-slug interfacial drag model were done. Then the newly developed code RELAP5/MOD3.2/TIGHT was used to analyze the LOCA transients of conceptually designed reactor cores with three types of tight lattice. The results showed that the peak cladding temperatures in the reflooding phase are much higher than that in the blow-down phase. Through comparison between the calculation results of LOCA transients of the three types of tight lattice, it was found that with smaller pitch to diameter ratio, the peak cladding temperature was much higher. LPIS injection flow rate should be increased in order to keep the rod cladding temperature be within the LOCA criteria. Steam generation will prevent the coolant from flowing downstream of the channel in reactor cores with a very small flow area. From the reactor safety aspect and the economic aspect, we do not recommend that reactor cores be designed with p/d ratio less than 1.10.     

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.     

Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK-1500 reactor main circulation circuit model

Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines.     

Validation of coupled neutronic/thermal-hydraulic code RELAP5-3D for RBMK-1500 reactor analysis application

This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.     

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.     

Experimental validation of the TASS/SMR code for an integral type pressurized water reactor

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.     

Modelling of low-pressure subcooled flow boiling using the RELAP5 code

A new model for upward vertical subcooled flow boiling at low pressure has been proposed. The model considers the most relevant closure relationships of one-dimensional thermal-hydraulic codes that are important for accurate prediction of vapour contents in the channel: wall evaporation model, condensation model, flow regime transition criterion and drift-flux model. The new model was incorporated in the current version of the RELAP5 code, MOD3.2.2 Gamma. The modified code was validated against a number of published low-pressure subcooled boiling experiments, and in contrast to the current code, shows good agreement with experimental data. The presented analysis also leads to a better understanding of the basic mechanisms of subcooled flow boiling at low pressure.     

Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code

Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid-steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is performed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data.     

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.     

On the use of RELAP5-3D as a subchannel analysis code

The capabilities of the RELAP5-3D code to perform subchannel analyses in sodium-cooled fuel assemblies were evaluated. The motivation was the desire to analyze fuel assemblies with traditional (solid pins) as well as non-traditional (e.g., annular pins with internal cooling, bottle-shape) geometries. Since no current subchannel codes can handle such fuel assembly designs, a new flexible RELAP5-based subchannel model was developed. It was shown that subchannel analysis of sodium-cooled fuel assemblies is indeed possible through the use of control variables in RELAP5. The subchannel model performance was then verified and validated in code-to-code and code-to-experiment analyses, respectively. First, the model was compared to the SUPERENERGY II code for solid fuel pins in a conventional hexagonal lattice. It was shown that the temperature predictions from the two codes agreed within 2% (<3.5 °C). Second, the model was applied to the Oak Ridge 19-pin test, and it was found that the measured outlet temperature distribution could be predicted with a maximum error of 8% (<7 °C). Furthermore, the use of semicircular ribs on the duct wall to flatten the temperature distribution in a traditional hexagonal assembly was explored by means of the newly developed RELAP5-3D subchannel model; the results are reported here as an example of the model capabilities.     

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.     

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.     

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.     

Dynamic analysis of a pressurized water reactor

An apparent malfunction in a pressurized water reactor system has been investigated using fluctuation analysis. Both frequency-domain and time-domain analyses have been used and the results obtained by the two methods have been compared. The recording was performed by a relatively simple, cheap system giving high recording precision and the analyses were performed on an IBM 370 digital computer. It is shown that, while considerable information can be derived from frequency-domain analyses, a misinterpretation can occur in some cases. Time-domain correlation, as normally performed, was not very informative. However, time-domain correlation on bandwidth-limited time-series proved to be very valuable and could remove the misinterpretation of the frequency-domain analyses. The bandwidth limitation was performed by digital filters.     

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is “Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)”. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed&Bleed procedure initiation. For the purpose of this, operator action with “Reactor vessel off-gas valve – 0.032 m” opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.     

Identification of pressurized water reactor transient using template matching

Template matching, which is a pattern recognition method, was adopted to identify the transient in a pressurized water reactor (PWR). The transient data were generated using a plant simulation code, PCTran-PWR, and transformed into a feature vector sequence (FVS). The data set contained such FVS as the reference transients. To compare the FVS of the test transient and reference transient, a cost function was defined and dynamic programming was applied to obtain the minimum cost function value, which would indicate the degree of matching between the two transients. Considering the discrepancy between the real plant and the model to generate the transient data, the same test and reference transient may not be matched exactly. A dynamic threshold value was designed to determine if the test transient matched the reference transient of the data set. Experiments were performed and the results showed that the method was successful.     

Assessment of RELAP5/MOD3.2.2 Gamma against ABB Atom 3×3-Rod Bundle Reflooding Tests

The results of the ABB Atom 3×3-Rod Bundle Reflooding Tests were used for assessment of the reflooding model used in RELAP5/MOD3.2.2 Gamma version. The assessment calculations were performed using the default calculation model options implemented in the code.The tests were performed to investigate the effects of different spacer grid designs on heat transfer during the reflooding period of a pressurized water reactor loss-of-coolant accident (LOCA). The tests were conducted under low-pressure and low-flow (LPLF) conditions using a PWR-type 3×3-rod bundle with full-length indirectly electrically heated, stepped cosine axial power-shaped heater rods. Three different spacer grid configurations were studied: spacer grids without mixing vanes, mixing vane spacer grids, and mixing vane spacer grids together with intermediate flow mixers (IFM).A total of 36 tests with different spacer grid configurations were calculated. For two selected basic tests with non-mixing spacer grids an extended comparison of calculated and measured parameters is presented. The comparison of the predicted and measured maximal cladding temperatures and quench times, which are the most important parameters in licensing calculations, is presented for all the performed tests.The assessment calculations were preceded by nodalization, time step, and moving mesh studies.The RELAP5/MOD3.2.2 Gamma code was found to still have several deficiencies in the reflood model. The calculation results show a satisfactory agreement with experimental inner peak cladding temperature, however the predicted temperature turn-around times and quench times are significantly too short. The results also show a significant over-prediction of the reflood heat transfer and the vapour temperatures. The void profile downstream the quench front is not correctly predicted either. Finally, the present reflood model does not properly reflect the effects of spacer grids on the reflood heat transfer.In spite of these deficiencies the improvements incorporated into RELAP5/MOD3.2 by the Paul Scherrer Institute (PSI) eliminated the unphysical behaviors such as continuous cooling without clear turn-around temperature and no visible quenching phenomena, which were shown in the reflood calculations by means of the RELAP5/MOD3.1 code.