DEVELOPMENT OF A BEST-ESTIMATE THERMAL HYDRAULICS MODEL OF THE HPR-1000 NPP FOR DEVELOPING/VERIFYING EOP
D. De Luca1, V. Parrinello1, S. Huang2, M. Cherubini1, A. Petruzzi1 and C. Yang2
1Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy
2China Nuclear Power Engineering Company, Beijing 100840, China
BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018
Abstract — Emergency Operating Procedures (EOPs) are essential for maintaining fundamental safety functions and preventing core damage during design basis accidents and beyond design basis accidents in a nuclear power plant. For the development and maintenance of EOPs and accident management procedures, best estimate codes together with realistic assumptions should be used. In this activity, a best estimate model of the HPR-1000 NPP has been developed and the preliminary simulations of four selected accident scenario (i.e. LOFW, MBLOCA, LBLOCA and SGTR) have been performed in order to investigate the system behaviour. Among these, the LOFW accident scenario with the feed and bleed procedures to remove the core decay heat is discussed in this paper. From the analysis of the results it can be seen that the effectiveness of the procedures mostly depends on the starting time of the operator actions and on the number of the PRZ safety valves that are request to operate. However, the present work has to be considered as a part of a more general framework of activities whose final goal is the development of a best estimate evaluation model of the selected NPP to be used for performing the accident safety analysis (including independent safety analysis activities to support the licensing) and the development of Emergency Operational Procedure (EOP). In order to reach this final goal, several additional activities must still be done, such as the validation of the evaluation model and the quantification of the uncertainties.
VALIDATION OF MULTI-PHYSICS SIMULATION TOOLS USING FUEL RAMP TEST IN R2 REACTOR: THE MPCMIV BENCHMARK
A. Petruzzi1, D. De Luca1, J. Karlsson2, T. Valentine3
1Nuclear and INdustrial Engineering - NINE, Via della Chiesa XXXII 759, Lucca, Italy
2Studsvik Nuclear AB, SE-611 82 Nyköping, Sweden
3Oak Ridge National Laboratory, Oak Ridge, TN, USA
BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018
Abstract — High-fidelity, multi-physics Modeling and Simulation (M&S) tools are being developed and utilized for a variety of applications in nuclear science and technology and show great promise in their abilities to reproduce observed phenomena for many applications. Even with the increasing fidelity and sophistication of coupled multi-physics M&S tools, the underpinning models and data still need to be validated against experiments that may require a more complex array of validation data because of the great breadth of the time, energy and spatial domains of the physical phenomena that are being simulated. The Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation (MPEBV) of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) was formed to address the challenges with the validation of such tools. In this framework, NINE and STUDSVIK promoted a benchmark titled Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV). The MPCMIV exercise is based upon cold ramp tests conducted at the Studsvik R2 reactor test loop that requires coupling of reactor physics, thermal hydraulics, and fuel performance phenomena. The aim of the experiment was to investigate the fuel response at cold criticality conditions to examine whether or not the potential fuel failure mechanisms might differ at temperatures below 100 °C than at normal operating conditions. The validation exercise is structured into four tiers in order to maximize participation by various groups. The first tier is targeted for novel M&S tools that have the capability to model the 3D heterogeneous model for both the reactor core domain and the fuel rod domain; the second tier involves the use of a simplified model for novel M&S tools that utilizes boundary conditions for the reactor physics models of the R2 reactor core; the third tier involves the same simplified model of tier 2 but allows for the use of traditional M&S tools; and the four tier is target for the application of only fuel performance tools. For each tier, the MPCMIV exercise is structured into four phases: the model development phase, the pre-qualification phase, the blind simulation phase, and the post-test phase. For each of these phases the participants will establish their validation requirements and assumptions that are made whether it be for steady state models or transient models.
MCNP6 UNCERTAINTY QUANTIFICATION APPLIED TO UAM-EXSERCISE I-1
L. Lampunio1, V. Giusti2, V. Parrinello1 and A. Petruzzi1
1Nuclear and INdustrial Engineering - NINE, Via della Chiesa XXXII 759, Lucca, Italy
2University of Pisa, Largo Lucio Lazzarino 2, Pisa (PI), Italy
BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018
Abstract — Uncertainty quantification has proved itself to be a more and more important aspect of the nuclear society for best estimate predictions to be provided with their confidence bounds. Besides, the ability to compute the multiplication factor sensitivity coefficients for uncertainty estimation is also useful for code validation, the development of benchmark suites applicable to specific sets of applications and the design of critical (integral) experiments. In the context of the OECD/NEA project on Benchmark for Uncertainty Analysis in Modelling (UAM), the MCNP 6.1 code has been used for sensitivity and uncertainty analysis in criticality calculations for the Exercise 1 “Cell physics”, Phase I. MCNP 6.1 is the first version of MCNP with the possibility to calculate the sensitivity coefficients of the multiplication factor k for nuclear data. The methods employed are based upon linear-perturbation theory using adjoint weighting, performed in a single forward calculation by means of the Iterated Fission Probability method. The test case under consideration is the Three Mile Island Pressurized Water Reactor fuel pin in Hot Zero Power and Hot Full Power conditions. The results of the MCNP 6.1 code are presented and compared with the ones already available obtained using Serpent and SCALE 6.0 codes. In particular, the results used for the present comparison had been produced by the TSUNAMI-1D module of SCALE and Serpent version 2.1.22. It was found a good agreement between the aforementioned codes in k uncertainty, though there are differences when the sensitivity profiles are analysed. The discrepancy obtained with the three codes in the k values are also presented.
SCCREDD METHODOLOGY FOR V&V: APPLICATION TO ATLAS A5.1 TEST BENCHMARK
V. Parrinello and M. Cherubini
Nuclear and INdustrial Engineering - NINE, Via della Chiesa XXXII 759, Lucca, Italy
BEPU 2018 - ANS Best Estimate Plus Uncertainty International Conference
Lucca, Italy, May 13-18, 2018
Abstract — The availability of an experimental and calculated qualified database is of outmost importance for the validation and qualification of codes in the framework of a BEPU application. The Standardized Consolidated Calculated and Reference Experimental Database (SCCRED) constitutes an approach envisaged also by the International Atomic Energy Agency to set up a qualified experimental and calculated database for verification and validation (V&V) purposes of computational tools and uncertainty methodologies. Such a database can be used to demonstrate that the code results are reliable, to perform an independent code assessment and also as a basis for developing and validating an uncertainty methodology. The present paper deals with the partial and simplified application of the SCCRED qualification procedure to the thermal-hydraulic analysis carried out for the ATLAS A5.1 post-test calculation in the framework of the . The adopted system thermal-hydraulic code to perform the analysis is RELAP5Mod3.3. The ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is a full pressure, full temperature integral test facility operated by KAERI (Korean Atomic Energy Research Institution). The ATLAS A5.1 experiment is part of the OECD-ATLAS Project and was performed targeting a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action by means of secondary side depressurization. The aims of the experiment are to demonstrate the progression of the cold leg line break accident and contribute to the understanding of the behavior of nuclear reactor systems and to the assessment and improvement of the existing best estimate codes. The developed nodalization has been qualified following the qualification procedure embedded into the SCCRED methodology since it has a geometrical fidelity with the ATLAS facility, it reproduces the measured nominal steady state condition of the system and shows a satisfactory behavior in time dependent conditions..
APPLICATION OF THE TRANSURANUS CODE TO HIGH BURN-UP LOCA TEST IN VIEW OF 10 CFR 50.46c
M. Cherubini, L. Lampunio
Nuclear and INdustrial Engineering (NINE), Borgo Giannotti, Lucca, Italy
TopFuel 2018
Prague, Czech Republic, 30 September - 04 October 2018
Abstract — Nowadays, the implementation of new cladding materials and higher rod burn-ups have led to the necessity of re-examining the LOCA safety criteria and verifying their validity regarding the importance of hydrogen content on cladding embrittlement. U.S – NRC 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light water nuclear power reactors,” is currently under revision to account for the influence of hydrogen on cladding embrittlement under LOCA conditions. In this work, the tests 191 and 192 conducted at the Hot Cell Laboratory of Studsvik Nuclear AB were simulated with TRANSURANUS code. Two base irradiation simulations were performed for both the father rod and the fuel rodlet to achieve consistent results. Consequently, to deal with the LOCA test, the restart option of TRANSURANUS was used to set the appropriate boundary conditions. The simulated data proved to be in agreement the experimental values for both phases of the exercise.