OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions

OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes) can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters). One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT) conducted by the Nuclear Power Engineering Corporation (NUPEC) in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWRtype fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermalhydraulic conditions, both in steady-state and transient conditions

Modeling the Heat Transfer of Helical Coil Tubes Steam Generator in SMR by RELAP5 Code and Validation

OSU MASLWR integral test facility, owned and operated at Oregon State University, represents a SMR cooled and moderated by light water. The main features of the experimental facility are based on the preliminary design of the Multi-Application Small Light Water Reactor carried out at Idaho National Labs. This reactor relies on natural circulation and on passive and inherently safe engineering safety features. It is featured by a helical-coil steam generator. The primary system and the containment are inherently coupled, in some anticipated operation occurrences and postulated accident conditions. Reliable simulations of such system requires the demonstration that a thermal-hydraulic system code, such as RELAP5 in this paper, can predict or, at least, bound some phenomena, which are outside its standard area of application. Challenging phenomena for the code simulation are: heat exchange for helical-coil SG and two phase instabilities in parallel tubes (SG secondary side); mixing and thermal stratification in a pool system; the condensation on the free surface in presence of noncodensable gas; chocked flow, the coupling primary system and containment. This paper presents also the performance of RELAP5 code in simulating the heat transfer in tube bundle with crossflow and the validation performed using two experiments performed in OSU MASLWR integral test facility. They are: 1) a natural circulation experiment aimed at characterizing the system performances at different power levels, and 2) a total loss of feedwater flow postulated accident scenario. The activity has been performed in the framework of an International Collaborative Standard Problem under the auspices of the IAEA

Predictability of CNEA PHWR MOX Experiments by mean of TRANSURANUS Code, from the IFPE Database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. The comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident conditions are part of the defense in depth concept. In this connection, OECD NEA sets up the public domain database on nuclear fuel performance experiments – International Fuel Performance Experiments (IFPE) database, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for models development and codes validation. The CNEA’s PHWR MOX Experiment belong to this database. This experiment was carried out in the High Flux Reactor (HFR) of Petten, Holland. It involves six MOX rods prepared and controlled in the CNEA’s “Alpha” Facility (Argentina). The objective of the experiment was to verify the fabrication processes and the study of the fuel behavior with respect to cladding failure due to Stress Corrosion Cracking (SCC) under Pellet Cladding Interaction (PCI) conditions. These rods were irradiated from 0 until 15 MWd/kgU. The code TRANSURANUS version “v1m1j09” is assessed against the database CNEA PHWR MOX in order to verify the capability of the code in predicting the cladding failures due to SCC. Comparisons with the experimental data and the results obtained with BACO code developed by “CNEA” group of Bariloche are presented. Sensitivity calculations are also performed for supporting the analyses of the results, improving the level of understanding of the code capabilities. The main conclusion is that the clad failure propensity of the rods belonging to the PHWR CNEA MOX experiment is conservatively assessed

Investigation of PCI phenomenon in PWR fuel with Transuranus code

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zircaloy cladding for model development and code validation. This paper describes the application of the TRANSURANUS code against power ramp tests executed in the research reactor R2 at Studsvik, Sweden. The datasets are part of the International Fuel Performance Experiments (IFPE) database. The experiments address the behavior of PWR fuel rods, including preceding base irradiation, during the over-power ramping. The burnup values range between 28 and 45 MWd/kgU. Pre-, during-, and post- irradiation, non destructive and destructive examinations were executed, in order to determine and understand the behavior of the fuel rods, but also to provide suitable data, useful for code validation. The experimental data were used for assessing the TRANSURANUS capabilities in predicting the phenomenon of the pellet clad interaction (PCI). The objective of the activity has been fulfilled developing input decks suitable for the assessment of TRANSURANUS code version “v1m1j08”. The current paper reports the main outcome of the assessment of the calculations

Modeling Large Break-LOCA: In Reactor Fuel Bundle Materials Test MT-4 and MT-6A

When creating power from nuclear fission, the fuel matrix and its cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. In this connection, OECD/NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation (IFPE)”. This database includes the data set of the projects MT-4 and MT-6A analyzed in the current paper. The MT-4 test bundle simulated a 6x6 section of a 17x17 3% enriched, full-length non-irradiated PWR fuel assembly. There were 20 non-pressurized guard fuel rods to isolate the 12 central, pressurized tests rods; the four corner rods were deleted. In the MT-6A test, the 20 guard rods used in the previous tests were replaced with 9 pressurized thus, a total of 21 test rods were in MT-6A. Only limited destructive post irradiation examination was performed on these two tests. The objective of the activity is the validation of TRANSURANUS “v1m1j09” code in predicting fuel and cladding behavior under LOCA conditions using the experimental databases MT-4 and MT-6A. It is pursued assessing the capabilities of the code models in simulating the phenomena and parameters involved, such as: pressure trend in the fuel rod, cladding creep, ? to ?-phase phase transformation, oxidation, geometry changes and finally failure prediction. The analysis is aimed at having a comprehensive understanding of the applicability and limitations of the code in the conditions of the experiments. Finally, probabilistic calculations are performed to complete the analysis. The objective of the activity is fulfilled addressing the behavior of two equivalent full lengths fuel rods, one for each test., suitable for the assessment of TU code versions “v1m1j09”

Modeling of BWR Inter-Ramp Project experiments by means of TRANSURANUS code

During the normal operation of a Light Water Reactor (LWR), the fuel–cladding gap may close, as a result of several phenomena and processes, including different thermal expansion and swelling of both the fuel and the cladding (Pellet Cladding Interaction – PCI). In this equilibrium state, a significant increase of local power (like a transient power ramp in the order of 100 kW/m h), induces circumferential stresses in the cladding. In presence of corrosive fission products (i.e. iodine) and beyond specific stress threshold depending on the material, cracks typical of stress corrosion may appear and grow-up (stress corrosion cracking – SCC). These cracks may spread out from the cladding internal surface, causing the fuel failure. Investigations on fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. To address the issue of PCI/SCC, the ‘‘BWR Inter-Ramp Project’’ is investigated by means of TRANSURANUS code. The BWR Inter-Ramp Project is part of the OECD/NEA International Fuel Performance Experiments database. It provides the experimental data of 20 BWR fuel rods during power ramp. The objective was to establish the failure-safe operating limits of representative BWR fuel rods when subjected to power ramp tests after short to medium irradiation time. The burn-up ranges from 10 to 20 MWd/kgU. The aim of the activity is to compare, investigate and summarize the main outcomes achieved after the simulations of 20 Zircaloy-2 – UO2 rods of various types and designs by TRANSURANUS code. Focus is given to the main variables, which are involved or may influence the simulation of cladding failure and characterize the power ramp tests. Sensitivity analyses are carried out to address the relevance of the knowledge of the boundary conditions, as well as the impact of selected parameters and code options on the simulations. The analyses bring to the conclusion that the code may under-estimate the pellet gaseous swelling (and consequently the cladding failures), of six fuel rods that experienced low power cycles at 10 MWd/kgU. The failure propensity of the remaining fourteen rods is conservatively simulated

Capabilities of Transuranus Code in Simulating Power Ramp Tests from the IFPE Database

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors. The TRANSURANUS code consists of a clearly defined mechanical–mathematical framework into which physical models can easily be incorporated. The mechanical–mathematical concept consists of a superposition of a one-dimensional radial and axial description (the so called quasi two-dimensional or 1½-D model). The code was specifically designed for the analysis of a single cylindrical rod. In the current paper, the application of the TRANSURANUS code to the Studsvik BWR Inter-Ramp and PWR Super-Ramp Projects are presented. The activity has been performed in the framework of an agreement between JRC-ITU and the University of Pisa and also of the IAEA FUMEX III project. The objective of the activity is the assessment of the fission gas release model of TRANSURANUS code verison“v1m1j08”, against the above mentioned databases. It constitutes an independent verification of a new model that was implemented at ITU for dealing with release during rapid power changes . The dataset of the BWR Inter-Ramp and PWR Super-Ramp Projects are part of the International Fuel Performance Experiments (IFPE) database. The first addresses the behavior of twenty standard-type unpressurised BWR fuel rods, including preceding base irradiation, during the over-power ramping. Two different values of base irradiations burnup were adopted for the experimental database: about 10 and 20 MWd/kgU. The latter addresses the behavior of twenty-eight light water reactor fuel rods when subject to power ramps (twenty-six are modeled for the current activity), after base irradiation to high burnup (28 to 45 MWd/kgU). Pre-, during-, and post- irradiation, non destructive and destructive examinations were executed, in order to determine and understand the behavior of the fuel rods, but also to provide suitable data, useful for code validation. The experimental data were used for assessing the TRANSURANUS capabilities in predicting the fission gas release. Focus is given on the prediction of the different FGR model options available in the code. The objective of the activity has been fulfilled developing forty-six input decks suitable for the assessment of TRANSURANUS code version “v1m1j08” , . The assessment is focused on fission gas release models available in TRANSURANUS code version “v1m1j08”, with particular emphasis to the new “TFGR model” which is implemented for taking into account the events of rapid power variations. The current paper reports the main outcome of the assessment of the calculations. Conclusive remarks of the activity are provided in the last section