Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known subchannel code COBRA-TF, namely CTF, to the critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks

Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

Selected papers from OECD-NEA PSBT benchmark

Historically, the prediction of safety margins has been based on system level thermal-hydraulic calculations employing suitable empirical formulations for assembly specific geometries and fuel-element grid spacers. These works have assessed response, margins, and consequences for the system based on one-dimensional two-fluid or drift-flux type thermalhydraulics formulations with fuel-vendor specific hydraulic losses and heat transfer characteristics for various fuel assemblies, including the so-called hot channel. Analysis of the hot channel gives important information on flow rates, fuel element centerline temperature, fuel sheath temperature, and margin to the departure from nucleate boiling. Given the reliance of the above approaches on empirical formulations obtained from complex and often difficult experiments, there is significant interest in obtaining reliable and accurate results from computation tools which employ more fundamental empirical relationships which can be obtained from subsets of the domain or from other scaled experiments.

A Comparison of different Uncertainty Activation Cross-Section Data Libraries: Application to the Prediction Uncertainty in Tritium Production

T actitivity in LiPb LiPb mock-up material irradiated in Frascati: measurement and MCNP results

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

UPM Results BWR-PWR Exercise I-1 by MCNP5 code

The perturbation approach relies in principle on a unique “NJOY + MCNP5 + SUSD3D”calculation. The inputs are the geometry MCNP5 input file and an ENDF file containing covariances.

Burnup Uncertainty Analysis for PWR-HFP Exercise I-1

Burn-up credit analyses are based on depletion calculations that provide an accurate prediction of spent fuel isotopic contents, followed by criticality calculations to assess keff

Isotopic prediction calculation methodologies: application to Vandellos-II Reactor cycles 7-11

Determining as accurate as possible spent nuclear fuel isotopic content is gaining importance due to its safety and economic implications. Since nowadays higher burn ups are achievable through increasing initial enrichments, more efficient burn up strategies within the reactor cores and the extension of the irradiation periods, establishing and improving computation methodologies is mandatory in order to carry out reliable criticality and isotopic prediction calculations. Several codes (WIMSD5, SERPENT 1.1.7, SCALE 6.0, MONTEBURNS 2.0 and MCNP-ACAB) and methodologies are tested here and compared to consolidated benchmarks (OECD/NEA pin cell moderated with light water) with the purpose of validating them and reviewing the state of the isotopic prediction capabilities. These preliminary comparisons will suggest what can be generally expected of these codes when applied to real problems. In the present paper, SCALE 6.0 and MONTEBURNS 2.0 are used to model the same reported geometries, material compositions and burn up history of the Spanish Van de llós II reactor cycles 7-11 and to reproduce measured isotopies after irradiation and decay times. We analyze comparisons between measurements and each code results for several grades of geometrical modelization detail, using different libraries and cross-section treatment methodologies. The power and flux normalization method implemented in MONTEBURNS 2.0 is discussed and a new normalization strategy is developed to deal with the selected and similar problems, further options are included to reproduce temperature distributions of the materials within the fuel assemblies and it is introduced a new code to automate series of simulations and manage material information between them. In order to have a realistic confidence level in the prediction of spent fuel isotopic content, we have estimated uncertainties using our MCNP-ACAB system. This depletion code, which combines the neutron transport code MCNP and the inventory code ACAB, propagates the uncertainties in the nuclide inventory assessing the potential impact of uncertainties in the basic nuclear data: cross-section, decay data and fission yields

Análisis de Estimación Óptima del Accidente de Pérdida de Refrigerante por Rotura Pequeña de 1”en Cabeza de Vasija en un Reactor tipo PWR

En el año 2002 durante una inspección se localizó una importante corrosión en la cabeza de la vasija de Davis Besse NPP. Si no se hubiera producido esa detección temprana, la corrosión hubiera provocado una pequeña rotura en la cabeza de la vasija. La OECD/NEA consideró la importancia de simular esta secuencia en la instalación experimental ROSA, la cual fue reproducida posteriormente por grupos de investigación internacionales con varios códigos de planta. En este caso el código utilizado para la simulación de las secuencias experimentales es TRACE. Los resultados de este test experimental fueron muy analizados internacionalmente por la gran influencia que dos factores tenía sobre el resultado: las acciones del operador relativas a la despresurización y la detección del descubrimiento del núcleo por los termopares que se encuentran a su salida. El comienzo del inicio de la despresurización del secundario estaba basado en la determinación del descubrimiento del núcleo por la lectura de los temopares de salida del núcleo. En el experimento se registró un retraso importante en la determinación de ese descubrimiento, comenzando la despresurización excesivamente tarde y haciendo necesaria la desactivación de los calentadores que simulan el núcleo del reactor para evitar su daño. Dada las condiciones excesivamente conservadoras del test experimentale, como el fallo de los dos trenes de inyección de alta presión durante todo el transitorio, en las aplicaciones de los experimentos con modelo de Almaraz NPP, se ha optado por reproducir dicho accidente con condiciones más realistas, verificando el impacto en los resultados de la disponibilidad de los trenes de inyección de alta presión o los tiempos de las acciones manuales del operador, como factores más limitantes y estableciendo el diámetro de rotura en 1”

Análisis de acciones del operador en un SBLOCA de 1”en el fondo de la Vasija en un Reactor tipo PWR

La simulación de accidentes de rotura pequeña en el fondo de la vasija se aparta del convencional análisis de LOCA de rama fría, el más limitante en los análisis deterministas La rotura de una de las penetraciones de instrumentación de la vasija ha sido desestimada históricamente en los análisis de licencia y en los Análisis Probabilistas de Seguridad y por ello, hay una falta evidente de literatura para dicho análisis. En el año 2003 durante una inspección, se detectó una considerable corrosión en el fondo de la vasija de South Texas Project Unit I NPP. La evolución en el tiempo de dicha corrosión habría derivado en una pequeña rotura en el fondo de la vasija si su detección no se hubiera producido a tiempo. La OECD/NEA consideró la importancia de simular dicha secuencia en la instalación experimental ROSA, la cual fue reproducida posteriormente por grupos de investigación internacionales con varios códigos de planta. En este caso el código utilizado para la simulación de las secuencias experimentales es TRACE. Tanto en el experimento como en la simulación se observaron las dificultades de reinundar la vasija al tener la rotura en el fondo de la misma, haciendo clave la gestión del accidente por parte del operador. Dadas las condiciones excesivamente conservadoras del test experimental, como el fallo de los dos trenes de inyección de alta presión durante todo el transitorio, en las aplicaciones de los experimentos con modelo de Almaraz NPP, se ha optado por reproducir dicho accidente con condiciones más realistas, verificando el impacto en los resultados de la disponibilidad de los trenes de inyección de alta presión o los tiempos de las acciones manuales del operador, como factores más limitantes y estableciendo el diámetro de rotura en 1”

Development of mathematical models to elaborate strategies, select alternatives and development of plans for adaptation of communities to climate change in different geographical areas including costs to implement it.

There is evidence that the climate changes and that now, the change is influenced and accelerated by the CO2 augmentation in atmosphere due to combustion by humans. Such ?Climate change? is on the policy agenda at the global level, with the aim of understanding and reducing its causes and to mitigate its consequences. In most countries and international organisms UNO (e.g. Rio de Janeiro 1992), OECD, EC, etc . . . the efforts and debates have been directed to know the possible causes, to predict the future evolution of some variable conditioners, and trying to make studies to fight against the effects or to delay the negative evolution of such. The Protocol of Kyoto 1997 set international efforts about CO2 emissions, but it was partial and not followed e.g. by USA and China . . . , and in Durban 2011 the ineffectiveness of humanity on such global real challenges was set as evident. Among all that, the elaboration of a global model was not boarded that can help to choose the best alternative between the feasible ones, to elaborate the strategies and to evaluate the costs, and the authors propose to enter in that frame for study. As in all natural, technological and social changes, the best-prepared countries will have the best bear and the more rapid recover. In all the geographic areas the alternative will not be the same one, but the model must help us to make the appropriated decision. It is essential to know those areas that are more sensitive to the negative effects of climate change, the parameters to take into account for its evaluation, and comprehensive plans to deal with it. The objective of this paper is to elaborate a mathematical model support of decisions, which will allow to develop and to evaluate alternatives of adaptation to the climatic change of different communities in Europe and Latin-America, mainly in especially vulnerable areas to the climatic change, considering in them all the intervening factors. The models will consider criteria of physical type (meteorological, edaphic, water resources), of use of the ground (agriculturist, forest, mining, industrial, urban, tourist, cattle dealer), economic (income, costs, benefits, infrastructures), social (population), politician (implementation, legislation), educative (Educational programs, diffusion) and environmental, at the present moment and the future. The intention is to obtain tools for aiding to get a realistic position for these challenges, which are an important part of the future problems of humanity in next decades.

Development of mathematical models to elaborate strategies, to select alternatives and the development of plans for the adaptation of communities to climate change in different geographical areas including costs to implement it

Climate change is on the policy agenda at the global level, with the aim of understanding and reducing its causes and to mitigate its consequences. In most of the countries and international organisms UNO, OECD, EC, etc … the efforts and debates have been directed to know the possible causes, to predict the future evolution of some variable conditioners, and trying to make studies to fight against the effects or to delay the negative evolution of such. Nevertheless, the elaboration of a global model was not boarded that can help to choose the best alternative between the feasible ones, to elaborate the strategies and to evaluate the costs. As in all natural, technological and social changes, the best-prepared countries will have the best bear and the more rapid recover. In all the geographic areas the alternative will not be the same one, but the model should help us to make the appropriated decision. It is essential to know those areas that are more sensitive to the negative effects of climate change, the parameters to take into account for its evaluation, and comprehensive plans to deal with it. The objective of this paper is to elaborate a mathematical model support of decisions, that will allow to develop and to evaluate alternatives of adaptation to the climatic change of different communities in Europe and Latin-America, mainly, in vulnerable areas to the climatic change, considering in them all the intervening factors. The models will take into consideration criteria of physical type (meteorological, edaphic, water resources), of use of the ground (agriculturist, forest, mining, industrial, urban, tourist, cattle dealer), economic (income, costs, benefits, infrastructures), social (population), politician (implementation, legislation), educative (Educational programs, diffusion), sanitary and environmental, at the present moment and the future.

Presentation and discussion of the UAM/exercise I-1b: "Pin-Cell Burn-Up Benchmark" with the hybrid method

The aim of this work is to present the Exercise I-1b “pin-cell burn-up benchmark” proposed in the framework of OECD LWR UAM. Its objective is to address the uncertainty due to the basic nuclear data as well as the impact of processing the nuclear and covariance data in a pin-cell depletion calculation. Four different sensitivity/uncertainty propagation methodologies participate in this benchmark (GRS, NRG, UPM, and SNU&KAERI). The paper describes the main features of the UPM model (hybrid method) compared with other methodologies. The requested output provided by UPM is presented, and it is discussed regarding the results of other methodologies.

Concrete constitutive model, calibration and applications

A methodology has been developed for characterising the mechanical behaviour of concrete, based on the damaged plasticity model, enriched with a user subroutine (V)USDFLD in order to capture better the ductility of the material under moderate confining pressures. The model has been applied in the context of the international benchmark IRIS_2012, organised by the OECD/NEA/CSNI Nuclear Energy Agency, dealing with impacts of rigid and deformable missiles against reinforced concrete targets. A slightly modified version of the concrete damaged plasticity model was used to represent the concrete. The simulation results matched very well the observations made during the actual tests. Particularly successful predictions involved the energy spent by the rigid missile in perforating the target, the crushed length of the deformable missile, the crushed and cracked areas of the concrete target, and the values of the strains recorded at a number of locations in the concrete slab.

Propagation of Cross-Section Uncertainties in Criticality Calculations in the Framework of UAM-Phase I Using MCNPX-2.7e and SCALE-6.1

In the framework of the OECD/NEA project on Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation, and Safety Analysis of LWRs, several approaches and codes are being used to deal with the exercises proposed in Phase I, “Specifications and Support Data for Neutronics Cases.” At UPM, our research group treats these exercises with sensitivity calculations and the “sandwich formula” to propagate cross-section uncertainties. Two different codes are employed to calculate the sensitivity coefficients of to cross sections in criticality calculations: MCNPX-2.7e and SCALE-6.1. The former uses the Differential Operator Technique and the latter uses the Adjoint-Weighted Technique. In this paper, the main results for exercise I-2 “Lattice Physics” are presented for the criticality calculations of PWR. These criticality calculations are done for a TMI fuel assembly at four different states: HZP-Unrodded, HZP-Rodded, HFP-Unrodded, and HFP-Rodded. The results of the two different codes above are presented and compared. The comparison proves a good agreement between SCALE-6.1 and MCNPX-2.7e in uncertainty that comes from the sensitivity coefficients calculated by both codes. Differences are found when the sensitivity profiles are analysed, but they do not lead to differences in the uncertainty.