Davor Grgić

The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

Abstract A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock & Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark. Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference: 1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling 2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling) 3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code. The influence of PVM and of direct coupling is also discussed. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper. The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some “recriticality” or “return to power” whose magnitude is largely affected by boundary and initial conditions.

Potential advantages and disadvantages of sequentially building small nuclear units instead of a large nuclear plant

Abstract Renewal of nuclear power programs in countries with modest electricity consumptions and weak electrical grid interconnections has raised the question of optimal nuclear power plants sizes for such countries. The same question would be also valid for isolated or weakly connected regions within a large country. Building large size nuclear power plant could be prevented by technical or financial limits. Research programs have been initiated in the International Atomic Energy Agency and in the USA (within the framework of the Global Nuclear Energy Partnership (GNEP) program) with the aim to inspect under which circumstances small and medium reactors could be the preferred option compared to large nuclear plants. The economy of scale is a clear advantage of large plants. This paper compares, by using probabilistic methods, the net cash flow of large and medium size plants, taking as example a large nuclear plant (around 1200 MW) and four sequentially built smaller plants (300 MW). Potential advantages and disadvantageous of both options have been considered. Main advantages of the sequential construction of several identical small units could be the reduced investor risk and reduced investment costs due to the learning effect. This analysis is a part of studies for the Croatian power generating system development.

Analysis of future nuclear power plants competitive investment costs with stochastic methods

Abstract One of main issues in planning future power system expansion is forecast of expected electricity generating costs. Due to the fact that the forecast should be extended to plants lifetime it is best to consider and compare lifetime levelized cost of generated electricity for each of the candidate plants. The analysis is a part of studies for the future expansion of the Croatian power system, but is valid generally. The main competitors to nuclear power plants are combined cycle natural gas fired plants whose economy strongly depends upon gas cost. The expected increase of gas cost in following decades can substantially influence the competitiveness of nuclear plants versus gas fired plants. In addition, the costs of electricity produced by coal fired plants and wind electricity generators were also considered. Due to uncertainties of input parameters it is advantageous to use probabilistic instead of deterministic method of analysis. The paper considers the problem of the probabilistic distribution of nuclear power plant investment costs for which such plants could be competitive to gas fired plants for an estimated probable range of gas cost increase rates.

Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fur Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

SPES3 Facility RELAP5 Sensitivity Analyses on the Containment System for Design Review

An Italian MSE R&D programme on Nuclear Fission is funding, through ENEA, the design and testing of SPES3 facility at SIET, for IRIS reactor simulation. IRIS is a modular, medium size, advanced, integral PWR, developed by an international consortium of utilities, industries, research centres and universities. SPES3 simulates the primary, secondary and containment systems of IRIS, with 1:100 volume scale, full elevation and prototypical thermal-hydraulic conditions. The RELAP5 code was extensively used in support to the design of the facility to identify criticalities and weak points in the reactor simulation. FER, at Zagreb University, performed the IRIS reactor analyses with the RELAP5 and GOTHIC coupled codes. The comparison between IRIS and SPES3 simulation results led to a simulation-design feedback process with step-by-step modifications of the facility design, up to the final configuration. For this, a series of sensitivity cases was run to investigate specific aspects affecting the trend of the main parameters of the plant, as the containment pressure and EHRS removed power, to limit fuel clad temperature excursions during accidental transients. This paper summarizes the sensitivity analyses on the containment system that allowed to review the SPES3 facility design and confirm its capability to appropriately simulate the IRIS plant.

I 2 S-LWR Concept Update

Pressurized water reactor of integral configuration (iPWR) offers inherent safety features, such as the possibility to completely eliminate large-break LOCA and control rod ejection. However, integral configuration implemented using the current PWR technology leads to a larger reactor vessel, which in turn, due to the vessel manufacturability and transportability restrictions, limits the reactor power. It is reflected in the fact that there are many proposed iPWR SMR concepts, with power levels up to approximately 300 MWe, but not many iPWR concepts with power level corresponding to that of large traditional PWR NPPs (900 MWe or higher). While SMRs offer certain advantages, they also have specific challenges. Moreover, large energy markets tend to prefer NPPs with larger power. The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept is an integral PWR, of larger power level (1000 MWe), that at the same time features integral configurations, and inherent safety features typically found only in iPWR SMRs. This is achieved by employing novel, more compact, technologies that simultaneously enable integral configuration, large power, and acceptable size reactor vessel. This concept is being developed since 2013 through a DOE-supported Integrated Research Project (IRP) in Nuclear Engineering University Programs (NEUP). The project led by Georgia Tech includes thirteen other national and international organizations from academia (University of Michigan, University of Tennessee, University of Idaho, Virginia Tech, Florida Institute of Technology, Brigham Young University, Morehouse College, University of Cambridge, Politecnico di Milano, and University of Zagreb), industry (Westinghouse Electric Company and Southern Nuclear), and Idaho National Laboratory. This concept introduces and integrates several novel technologies, including high power density core, silicide fuel, fuel/cladding system with enhanced accident tolerance, and primary micro-channel heat exchangers integrated with flashing drums into innovative power conversion system. Many inherent safety features are implemented as well, based on all passive safety systems, enhancing its safety performance parameters. The concept aims to provide both the enhanced safety and economics and offers the next evolutionary step beyond the Generation III + systems. This paper presents some details on the concept design and its safety systems and features, together with an update of the project progress.

Xenon Correction in Homogenized Neutron Cross Sections

The calculations were performed at different power levels for selected IFBA and non-IFBA NPP Krsko fuel assemblies using 2D lattice codes (DRAGON, NEWT and FA2D) to show quantitative influence of xenon concentration on homogenized cross sections. In addition to usual influence on fast and, to larger extent, on thermal absorption cross sections, similar type of influence is found in case of fission cross sections, and to some extent in fast to thermal removal cross section. Based on relative change of cross sections from the values obtained during reference depletion, burnup and power level dependent correction values were calculated and included in separate cross section library. Limited dependence of correction values on IFBA presence and enrichment is found. When applied within nodal code the correction has potential to improve prediction of multiplication factors (boron concentration) and axial power distribution in Pressurized Water Reactor (PWR) reactors.Copyright

The SPES3 Facility for Testing an Integral Layout SMR: BDBE Simulation Analysis

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.Copyright

Precursor Based PTS Screening Methodology of the EOP Operator Actions for PWR Plant

The reactor pressure vessel integrity must be preserved throughout the plant operation as well as in the postulated accidental conditions. Operator actions undertaken to mitigate consequences of the accidents should not increase the probability of Pressurized Thermal Shock transients (PTS), transients with high pressure in reactor coolant system and interaction of cold safety injection water and hot vessel wall, which can damage reactor pressure vessel. In order to provide bounding analysis for the development of the Emergency Operating Procedures (EOP) background analysis, precursor based PTS screening methodology was used, according to determined criteria, to select the most severe PTS scenarios for two loop Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). In order to bound asymmetrical behavior of the two loop plant, selected transients were further analyzed with vertically split reactor vessel model and precursor based analysis was made for the results of the split reactor vessel analysis. The results of these analyses have been used as boundary conditions for the two-dimensional FEM reactor vessel integrity analysis.Copyright