Analysis of unmitigated large break loss of coolant accidents using MELCOR code

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation

Analyses of the OSU-MASLWR Experimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well

Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes

The analyses of postulated severe accident sequences play a key role for the international nuclear technical scientific community for the study of the effect of possible actions to prevent significant core degradation and mitigate source term release. To simulate the complexity of phenomena involved in a severe accident, computational tools, known as severe accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2 ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a generic western three-loops PWR-900 MWe, has been carried out with the aim to give some insights on the modelling capabilities of these tools and to characterize the differences in the calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark exercise. In the postulated transient, the unavailability of all active injection coolant systems has been considered and only the injection of accumulators has been assumed as accident mitigation strategy

Scaling Issues for the Experimental Characterization of Reactor Coolant System in Integral Test Facilities and Role of System Code as Extrapolation Tool

The phenomenological analyses and thermal hydraulic characterization of a nuclear reactor are the basis for its design and safety evaluation. In light of the impossibility and huge cost of performing meaningful experiments at full scale, scaled down experimental tests - Integral Effect Test (IET) and Separate Effect Test (SET) - are more feasible in developing “assessment database”. The data are useful in characterizing the prototype design and in the validation of computational tools for safety analysis. The analyses of system behaviors including component interactions in the Reactor Coolant System (RCS), the Containment System (PCV) and the RCS/PCV coupled system have been extensively investigated using IETs in the past decades. Though several scaling methods, e.g. Linear, Power/Volume, Three level scaling, H2TS..., have been developed and applied in the IET and SET design, a direct extrapolation of the data to the prototype, i.e. the scalability, is in general not possible due to unavoidable scaling distortions. The scaling distortions are related to many factors, mainly the complex geometry, multiple component interactions and two phase thermal hydraulic phenomena in steady state and transient condition of a nuclear reactor. The complex nature of scaling a nuclear reactor requires a large number of scaling parameters to be simultaneously fulfilled. In addition, physical construction and funding constraints demand that a scaling compromise is inevitable. Therefore a scaling approach, e.g. time preserved/not preserved, full height/reduced height, full pressure/reduced pressure, full power/reduced power…, has to be adopted in accordance with the objective of the IET or SET. Together with the scaling analysis, Best Estimate (BE) thermal hydraulic system code has been used for supporting experiment activity (design facilities, interpretation of results, etc) and for extrapolating results to full scale prototype conditions. Since the closure laws in the system code are mainly based on scaled test data, the extrapolation of code results remains a challenging and open issue. Starting from a brief analysis of the main characteristics of IETs and SETFs, the main objective of this paper is to analyze some IET scaling approaches used to the simulation of RCS responses which characterize the main scaling limits. The scaling approaches and their constraints in ROSA-III, FIST and PIPER-ONE facility will be used to analyze their impact to the experimental prediction in Small Break LOCA counterpart tests. The liquid level behavior in the core and the core cladding temperature analysis are discussed used as judging criteria for the facilities scaling-up limits

Ingress of Coolant Event simulation with TRACE code with accuracy evaluation and coupled DAKOTA Uncertainty Analysis

Among the Postulated Initiating Events in nuclear fusion plants, the Ingress of Coolant Event (ICE) in the Plasma Chamber is one of the main safety issues. In the present paper, the best estimate thermal-hydraulic system code TRACE, developed by USNRC, has been adopted to study the ICE, and it has been qualified based on experimental results obtained in the Integrated ICE facility at JAERI. A nodalization has been developed in the SNAP environment/architecture, using also the TRACE 3D Vessel component where multidimensional phenomena could occur. The accuracy of the code calculation has been assessed both from a qualitative and quantitative point of view. In addition, an Uncertainty Analysis (UA), with the probabilistic method to propagate the input uncertainties, has been performed to characterize the dispersion of the results. The analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in the SNAP environment/architecture. Results show the adequacy of the 3D nodalization and the capability of the code to follow the transient evolution also at a very low pressure. Response correlations have been computed to characterize the correlation between the selected uncertain input parameters and the Plasma Chamber pressure

Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code

The Oregon State University (OSU) has constructed, under a U.S. Department of Energy grant, a scaled integral test facility to examine natural circulation phenomena characterizing the Multi-Application Small Light Water Reactor (MASLWR) design. The MASLWR is a small modular PWR relying on natural circulation during both steady-state and transient operation, which includes an integrated helical coil steam generator within the reactor pressure vessel. Testing has been conducted in order to assess the operation of the prototypical MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems performance. The experimental data produced are useful also for the assessment of the computational tools necessary for the operation, design and safety analysis of nuclear reactors. This report describes the assessment of TRACE code predictions, conducted under the NRC CAMP program, against the MASLWR tests OSU-MASLWR-001 and the OSU-MASLWR-002, respectively. This activity has been conducted in collaboration with the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA), the Department of Energy of the University of Palermo, the Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG) of University of Pisa, the Department of Nuclear Engineering and Radiation Health Physics at OSU and NuScale Power Inc. In particular the OSU-MASLWR-001 test, an inadvertent actuation of one submerged ADS valve, investigates the primary system to containment coupling under design basis accident conditions; the OSU-MASLWR-002 test, a natural circulation test, investigates the primary system flow rates and secondary side steam superheat for a variety of core power levels and feed water flow rates. The assessment against experimental data shows that the TRACE code predicts the main phenomena of interest of the selected tests reasonably well for most condition

SBO analysis of a generic PWR-900 with ASTEC and MELCOR codes

Abstract After the Fukushima accident, the interest of the public to nuclear safety has growth and the international technical nuclear community has increased his attention in the investigation and the characterization of Severe Accident (SA) scenarios. In order to simulate the different, complex and multi-physical phenomena involved in a SA, computational tools, known as SA codes, have been developed in the last decades. In order to give some insights on the modelling capabilities of these tools and the differences in the calculation results, also related to the user-effect, an analysis of an unmitigated Station Black Out (SBO) occurring in a generic Western three-loops PWR 900 MWe has been carried out by the authors in the framework of the NUGENIA TA-2 ASCOM project. The simulation results of ASTEC code (study carried out with ASTEC V2, IRSN all rights reserved, [2019]), developed by IRSN, and MELCOR 2.2 code, developed by SANDIA for USNRC, have been compared and analyzed. The SBO scenario considered takes into account the intervention of the accumulators as only accident mitigation strategy. Several figures of merits related to the thermal-hydraulic (e.g. primary pressure, cladding temperature, etc.) and to the core degradation (e.g. hydrogen production, etc.) have been considered to describe the accident evolution until the vessel failure, for the two codes comparison

Analyses of the OSU-MASLWRExperimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well

Current status of MELCOR 2.2 for fusion safety analyses

MELCOR is an integral code developed by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission (USNRC) to perform severe accident analyses of Light Water Reactors (LWR). More recently, MELCOR capabilities are being extended also to analyze non-LWR fission technologies. Within the European MELCOR User Group (EMUG), organized in the framework of USNRC Cooperative Severe Accident Research Program (CSARP), an activity on the evaluation of the applicability of MELCOR 2.2 for fusion safety analyses has been launched and it has been coordinated by ENEA. The aim of the activity was to identify the physical models to be possibly implemented in MELCOR 2.2 necessary for fusion safety analyses, and to check if those models are already available in MELCOR 1.8.6 for fusion version, developed by Idaho National Laboratory (INL). From this activity, a list of modeling needs emerged from the safety analyses of fusion-related installations have been identified and described. Then, the importance of the various needs, intended as the priority for model implementation in the MELCOR 2.2 code, has been evaluated according to the technical expert judgement of the authors. In the present paper, the identified modeling needs are discussed. The ultimate goal would be to propose to have a single integrated MELCOR 2.2 code release capable to cover both fission and fusion applications

Current status of Melcor 2.2 for fusion safety analyses

MELCOR is an integral code developed by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission (USNRC) to perform severe accident analyses of Light Water Reactors (LWR). More recently, MELCOR capabilities are being extended also to analyze non-LWR fission technologies. Within the European MELCOR User Group (EMUG), organized in the framework of the USNRC Cooperative Severe Accident Research Program (CSARP), an activity on the evaluation of the applicability of MELCOR 2.2 for fusion safety analyses has been launched and it has been coordinated by ENEA. The aim of the activity was to identify the physical models to be possibly implemented in MELCOR 2.2 necessary for fusion safety analyses, and to check if those models are already available in MELCOR 1.8.6 fusion version, developed by Idaho National Laboratory (INL). From this activity, a list of modeling needs that emerged from the safety analyses of fusion-related installations has been identified and described. Then, the importance of the various needs, intended as the priority for model implementation in the MELCOR 2.2 code, has been evaluated according to the technical expert judgment of the authors. In the present paper, the identified modeling needs are discussed. The ultimate goal would be to propose to have a single integrated MELCOR 2.2 code release capable to cover both fission and fusion applications