| 1. |
- Agung, Alexander, 1972, et al.
(författare)
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Development and Validation of Coupled PARCS/RELAP5 Model for Forsmark NPP at Uprated Power
- 2014
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Ingår i: The 10th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-10), Okinawa, Japan, December 14-18, 2014. ; , s. 1-10
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Konferensbidrag (refereegranskat)abstract
- ABSTRACTThis paper gives an account of the development and validation of an up-to-date coupled neutronic/thermal-hydraulic model for the Swedish Forsmark boiling water reactor. The model will be used for analyses of the consequences of the planned power uprate from 2928 MWth to 3253 MWth.At first, the development of the PARCS and RELAP5 models are presented. On the neutronic side, cross-sections data was generated, allowing feeding PARCS with realistic data. This step was performed by converting the library data file from the power plant using the in-house cross-section interface code. The dependence of the material properties on history effects, burnup, and instantaneous conditions was accounted for, and the full heterogeneity of the core was thus taken into account. Each of the 676 fuel assemblies was modeled individually, while the 161 control rods were grouped into 6 different types. On the thermal-hydraulic side, the model consists of a model for the feedwater system, a model for the reactor vessel that include a model for the core channels, and a model for each of the four steam lines. The fuel assemblies were modeled as twelve flow channels in the core region. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.The validation efforts were focusing on benchmarking the code capabilities against measured plant data, both under steady-state and transient conditions. The PARCS standalone model was validated against traversing in-core probe (TIP) measurements, taken at different burnup level with operating power varies from 108% (nominal level) to 120% (uprated level). The coupled PARCS/RELAP5 model was validated against an operational transient. For this validation task, the transient chosen was a turbine trip test, which was performed on May 6, 2013.Comparisons between calculated and measured parameters demonstrate that the coupled model was able to correctly represent the steady-state conditions of the plant. The validation of the coupled model against measured transient plant data was then performed. It has been demonstrated that the coupled model is able to capture the main features of the transient with a sufficient level of accuracy
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| 2. |
- Agung, Alexander, 1972, et al.
(författare)
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The Role of Swedish Universities in Supporting SSM Activities in the Field of Deterministic Safety Analysis
- 2012
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Ingår i: Prosiding Seminar Keselamatan Nuklir 2012, 4 July 2012, Jakarta. - 1412-3258. ; , s. 1-16
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Konferensbidrag (refereegranskat)abstract
- Research within nuclear safety and radiation protection is necessary in order to maintain the high level of competence required by an expert authority. In the field of reactor safety research, SSM’s goals are to support regulation and contribute to national competence in the area of nuclear safety. A technical support organization on deterministic safety analysis (TSO-DSA) has been set up to help SSM in fulfilling these goals. The TSO-DSA function was then established by SSM at two nuclear universities, i.e. Royal Institute of Technology (KTH) in Stockholm and Chalmers University of Technology in Gothenburg. Activities related to this function have been performed, emphasizing the use of best-estimate coupled codes (i.e. PARCS/RELAP5 and PARCS/TRACE) for the analyse . The activities performed by Chalmers are reported in this paper as examples. The on-going activities give a good example on how the safety authority co-operates with universities. The use of coupled codes gives satisfactory results and in good agreements with measured data. Moreover, it may reveal some phenomena that are difficult to capture with stand-alone codes.
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| 3. |
- Agung, Alexander, 1972, et al.
(författare)
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Validation of PARCS/RELAP5 Coupled codes against a Load Rejection Transient at the Ringhals-3 NPP
- 2013
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493. ; 257, s. 31 - 44
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Tidskriftsartikel (refereegranskat)abstract
- This article gives an account of the development and validation of the PARCS/RELAP5 model of the Ringhals-3 unit against a load rejection transient, which occurred on November 28, 2010. The third unit of the Ringhals nuclear power plant comprises a 3-loop Westinghouse design pressurized water reactor on the Swedish West Coast.At first, the development of the PARCS and RELAP5 models are presented. On the neutronic side, a unique cross-section interface, allowing feeding PARCS with realistic data, was developed. The dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. On the thermal-hydraulic side, all the 157 fuel assemblies are modeled individually in the code input. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.Comparison between calculated and measured parameters demonstrates that the coupled model is able to correctly represent the steady-state conditions of the plant. The validation of the coupled model against measured transient plant data is then performed. It is demonstrated that the coupled model is able to catch the main features of the transient with a sufficient level of accuracy.
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| 4. |
- Banati, Jozsef, 1960, et al.
(författare)
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Analysis of a loss of feedwater case at the Ringhals-3 NPP using RELAP5/PARCS coupled codes
- 2008
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Ingår i: 16th International Conference on Nuclear Engineering, ICONE16 2008; Orlando, FL; United States; 11 May 2008 through 15 May 2008. - 9780791848159 ; 3, s. 135-144
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Konferensbidrag (refereegranskat)abstract
- This paper deals with the development and validation of a coupled RELAP5/PARCS model of the Swedish Ringhals-3 pressurized water reactor against a Loss of Feedwater transient, which occurred on August 16, 2005.At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modeled in individually in both codes. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, because of the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions.Capabilities of the RELAP5 code were more challenged in this transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analysis. The calculated values of the parameters show good agreement with the measured data.
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| 5. |
- Banati, Jozsef, 1960, et al.
(författare)
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Analysis of a loss of normal feedwater transient at the Ringhals-3 NPP using RELAP5/Mod.3.3
- 2010
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report gives an account on the development and validation of the RELAP5/Mod3.3 model of the Ringhals-3 pressurized water reactor against a Loss of Normal Feedwater Transient, which occurred on August 16, 2005. The 3rd unit of Ringhals Nuclear Power Plant comprises a 3-loops Westinghouse design pressurized water reactor on the Swedish West Coast.At first, the RELAP5 model is presented. All the 157 fuel assemblies are modeled individually in the code input. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops.The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, due to the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were challenged in this transient. The calculated values of the parameters show good agreement with the measured data.A parametric study was performed In order to evaluate the dependence of the steam generator level on the injected auxiliary feedwater flow. It indicated that the turbine driven auxiliary feedwater pump could possibly inject at a higher flowrate than its nominal value.The work was performed by the Department of Nuclear Engineering, Chalmers University of Technology in the framework of the Ringhals-3 power uprate project, supported by the Swedish Radiation Safety Authority (SSM). The ultimate goal of this project is to perform independent safety analyses of some limiting transients associated to the power uprate. The work carried out so far was targeted towards the development of state-of-the-art modelling capabilities for the Ringhals-3 unit.The present validational study is a Swedish contribution to the international Code Assessment and Maintenance Program (CAMP).
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| 6. |
- Banati, Jozsef, 1960, et al.
(författare)
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Development and validation of a coupled PARCS/RELAP5 model of the Ringhals-3 PWR
- 2007
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Ingår i: Proc. 15th Int. Conf. Nuclear Engineering (ICONE-15).
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Konferensbidrag (refereegranskat)abstract
- This paper deals with the development of a coupled PARCS/RELAP5 model of the Swedish Ringhals-3 pressurized water reactor. The stand-alone PARCS and RELAP5 models are first presented. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. On the thermal-hydraulic side, each of the 157 fuel assemblies is modelled. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.Validation of the model against measured plant data at steady-state conditions is then summarized. Comparisons between calculated/measured parameters demonstrate that the model is able to correctly represent steady-state conditions of the plant.Finally, the validation of the model against measured transient plant data is described. The transient chosen for this validation task was a load rejection (“house-load”) transient, which occurred on January 8, 2005.
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| 7. |
- Banati, Jozsef, 1960, et al.
(författare)
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Development of a coupled PARCS/RELAP5 model of the Ringhals-3 PWR
- 2006
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Ingår i: Proc. Int. Conf. Advances in Nuclear Analysis and Simulation (PHYSOR2006), Vancouver, British Columbia, Canada, September 10-14, 2006, American Nuclear Society. - 0894486977
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Konferensbidrag (refereegranskat)abstract
- This paper deals with the development of a coupled PARCS/RELAP5 model of the Swedish Ringhals-3 pressurized water reactor. The stand-alone PARCS and RELAP5 models are first presented. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. On the thermal-hydraulic side, each of the 157 fuel assemblies is modeled. The model is furthermore able to handle possible asymmetrical conditions of the flow field between the loops. The coupling between the two codes is then reported, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. Preliminary coupling tests for steady-state calculations were successfully performed and comparisons against plant measured data demonstrate a very good agreement of the model with the measurements. The coupled model will be later on used for analyzing the consequences of the power uprate planned for this reactor, via the simulation of a few limiting transients.
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| 8. |
- Banati, Jozsef, 1960, et al.
(författare)
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Validation of relap5/mod3.3 against a load step transient at Ringhals 4 u
- 2015
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Ingår i: 16th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) Chicago, IL, August 30 - September 4 2015. - 9781510811843 ; 6, s. 4548-4561
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Konferensbidrag (refereegranskat)abstract
- ABSTRACTThe Unit 4 of the Ringhals Nuclear Power Plant has recently undergone a large component replacement project with installation of new steam generators and a pressurizer, targeting a power uprate. A series of startup and maneuverability tests has been performed, mainly focusing on evaluation of the system responses for various perturbations. The subject of the present numerical analysis is a test with a ± 10 % steps applied in the load. The data collected during test provided a good opportunity for validation of the full plant model, which was prepared recently for the RELAP5/Mod3.3 Patch04 computer code, with incorporation of the new component models.The paper introduces the test procedure and shows an overview on the key parameters that are utilized as initial and boundary conditions. Strategies applied for achievement of steady-state conditions are addressed in the document. Furthermore, the paper summarizes the results of the validation study using the transient data to simulate the startup test. It has been proven that the stand-alone RELAP5 thermal-hydraulic model is capable of reproduction of the key features and events of the test. Sufficiently good agreement has been achieved between the measured and simulated thermal hydraulic parameters, already in its current stage of model development. On the basis of successful verification at the original power, it is expectable that the new Ringhals 4 model will be able to predict the fluid conditions in other types of transients, even at uprated conditions.
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| 9. |
- Banati, Jozsef, 1960, et al.
(författare)
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Validation of RELAP5/Mod3.3 Against the PACTEL SBL-50 Benchmark Transient
- 2013
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Ingår i: The 15th International Topical Meeting on Nuclear Reactor Thermal - Hydraulics, NURETH-15, Pisa, Italy, May 12-17, 2013. ; , s. 1-12
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Konferensbidrag (refereegranskat)abstract
- ABSTRACTThe PWR PACTEL test facility has recently been designed to support the safety studies of EPR type nuclear reactor thermal-hydraulics. The facility is located at the Lappeenranta University of Technology (LUT) in Finland. It is essentially important to understand the system behavior under natural circulation conditions during a Loss of Coolant Accident (LOCA). With this objective in mind, an international benchmark transient was conducted at the LUT in 2010-2011. The SBL-50 test was a SB-LOCA with a 1 mm break in the cold leg. The continuous inventory loss led to core dry-out. This project gave unique opportunities for several organizations to build and validate their models for the PWR PACTEL, as well as to simulate the transient by using various computer codes. Chalmers University of Technology participated with RELAP5/Mod3.3 calculations both in the pre-test and post-test phases of the project. The pre-test simulation included a simplified steam generator model, with the description of the heat exchange by a single characteristic U-tube. This coarse nodalization resulted in reasonably good agreement with the measured data. As the test data became known in the post-test, minor modifications contributed to achievement of better results. The changes were related to the upper plenum nodalization and the critical discharge flow parameters at the break assembly. Even if the post-test model provided better agreement for most of the parameters, it still had difficulties to predict the temperatures in the longest tubes in the steam generators (SGs). A careful examination of the measured data indicated that discrepancies might originate from a flow reversal in the SG primary side. Thus, an advanced SG model was prepared with application of a multi-channel system. Altogether 51 heat exchanger tubes are arranged into 5 bundles in the PWR PACTEL SG, according to 5 different lengths. These bundles were individually modeled in the refined input. The most recent multi-channel SG model has not only confirmed the presence of reverse flow and internal circulation in the SG primary, but it has also contributed to a much better prediction of the fluid temperature distribution.
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| 10. |
- Banati, Jozsef, 1960, et al.
(författare)
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Validation of the RELAP5 Model of Ringhals 4 against a Load Step Test at Uprated Power
- 2015
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- ABSTRACTAfter finishing the large component replacement project of the steam generators and the pressurizer, the Unit 4 of the Ringhals Nuclear Power Plant was operated in test mode between 2011 and 2014. In order to reach the ultimate goal of power uprate, and obtain a license for operation at the new nominal conditions, a series of tests had to be accomplished.The R4-QP-101 maneuverability test was performed on 2015-03-03. The test was focusing on evaluation of the system responses for ± 10 % perturbances in the load. The data collected during test provided a good opportunity for verification of the full plant model, which is being developed for the RELAP5/Mod3.3 Patch04 computer code, with incorporation of the new component models.The present Scientific Report introduces the test procedure and shows an overview on the key parameters that are utilized. There is a brief summary given on the applicability field of the computer code. Preparation of the input deck and the nodalization of the primary and secondary sides are touched upon.Strategies applied for achievement of steady-state conditions are addressed in the document. The steady-state results are presented in plotted format, demonstrating how the control system brought the entire unit to stable conditions. The calculated steady-state parameters were very close to the measured plant data, before the transient initiation.A chapter summarizes the results of the validation study using the transient data to simulate the startup test. Quantification of simulation accuracy has proven that the stand-alone RELAP5 thermal-hydraulic model is capable of reproduction of the key features and events of the test. Sufficiently good agreement between the measured and simulated data resulted in a successful verification of the plant model. The Ringhals 4 model is suitable for analysis of other types of transients already in its current state of development. Nevertheless, further refinement in the input is planned as soon as new test data will be available.Keywords: Ringhals 4, Load Step, RELAP5, code validation
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