Benchmark Analysis of EBR-II Protected Loss-of-Flow Transient

Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA). The objective of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR). The Experimental Breeder Reactor II (EBR-II) plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S. Department of Energy at the Argonne-West site. In the frame of this project, benchmark analysis of one of the EBR-II shutdown heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed at the Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG) in Pisa, Italy. The aim of this paper is to present modeling of EBR-II reactor design using RELAP-3D, and to present results of the transient analysis of SHRT-17. Complete nodalization of the reactor was made from the beginning. Model is divided in primary side that contains core, pumps, reactor pool and, for this kind of reactor specific, Z pipe, and intermediate side that contains Intermediate Heat Exchanger (IHX). Core was modeled with 82 channels that represent all fuel assemblies, and 14 channels for reflector and blanket assemblies. After achievement of acceptable steady-state results, transient analysis was performed. Starting from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was scrammed to simulate a protected loss-of-flow accident. In addition, the primary system auxiliary coolant pump, that normally had an emergency battery power supply, was turned off. Despite early rise of the temperature in the reactor, the natural circulation characteristics managed to keep it at acceptable levels and cooled the reactor down safely at decay heat power levels. Thermal-hydraulics characteristics and plant behavior was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL

International training Program: 3D S.UN.COP – Scaling, Uncertainty and 3D thermal-hydraulics/neutron-kinetics CouPled Codes Seminar

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A sixth seminar will be organized in 2007 at the Texas A&M University involving more than 30 scientists between lecturers and code developers

International Training Program in Support of Safety Analysis: 3D S.UN.COP - Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron kinetics Coupled Codes Seminars

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Nine seminars have been held at University of Pisa (two in 2004), at The Pennsylvania State University (2004), at the University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January- February 2006), in Buenos Aires, Argentina (October 2006), requested by Autoridad Regulatoria Nuclear (ARN), Nucleoelectrica Argentina S.A (NA-SA) and Comisión Nacional de Energía Atómica (CNEA), at the College Station, Texas A&M, (January-February 2007), in Hamilton and Niagara Falls, Ontario (October 2007) requested by Atomic Energy Canada Limited (AECL), Canadian Nuclear Society (CNS) and Canadian Nuclear Safety Commission (CNSC), in Petten, The Netherlands (October 2008) in cooperation with the Institute of Energy of the Joint Research Center of the European Commission (IE-JRC-EC). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2008 at IE-JRC was successfully held with the attendance of 35 participants coming from more than 10 countries and 20 different institutions (universities, vendors and national laboratories). More than 30 scientists (coming from more than 10 countries and 20 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A tenth seminar will be held (October 2009) at the Royal Institute of Technology (KTH) in Amsterdam (Sweden), involving more than 30 scientists between lecturers and code developers (http://dimnp.ing.unipi.it/3dsuncop/2009/index.html)

The ‘Code-User Effect’ and the International Training Program in Support of Safety Analysis

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Seven seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006), in Buenos Aires, Argentina (October 2006, requested by ARN -Autoridad Regulatoria Nuclear- NA-SA -Nucleoelectrica Argentina S.A- and CNEA -Comisión Nacional de Energía Atómica) and at Texas A&M University (January-February 2007). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2007 at Texas A&M. (http://dimnp.ing.unipi.it/3dsuncop/2007) was successfully held with the attendance of 26 participants coming from 12 countries and 17 different institutions (universities, vendors, national laboratories and regulatory bodies). About 30 scientists (coming from 11 countries and 19 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate was released to participants that successfully solved the assigned problems

Benchmark Analysis of EBR‐II Shutdown Heat Removal Test SHRT‐17

Coordinated Research Project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT) was established by International Atomic Energy Agency (IAEA). The aim of the project is to support and to improve validation of simulation tools and projects for Sodium-cooled Fast Reactors (SFR). The Experimental Breeder Reactor II (EBR-II) plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S. Department of Energy at the Argonne-West site. In the frame of this project, benchmark analysis of one of the EBR-II shutdown heat removal tests, protected loss-of-flow transient (SHRT-17), has been performed at the Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG) in Pisa, Italy. The aim of this paper is to present modelling of EBR-II reactor design using RELAP-3D, and to present results of the transient analysis of SHRT-17. Complete nodalization of the reactor was made from the beginning. Model is divided in primary side that contains core, pumps, reactor pool and, for this kind of reactor specific, Z pipe, and intermediate side that contains Intermediate Heat Exchanger (IHX). After achievement of acceptable steady-state results, transient analysis was performed. Starting from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was scrammed to simulate a protected loss-of-flow accident. In addition, the primary system auxiliary coolant pump that normally had an emergency battery power supply was turned off. Despite early rise of the temperature in the reactor, the natural circulation characteristics managed to keep it at acceptable leveles and cooled the reactor down safely at decay heat power levels. Thermal-hydraulics characteristics and plant behaviour was focused on prediction of natural convection cooling by evaluating the reactor core flow and temperatures and their comparison with experimental data that were provided by ANL. The plan is to improve the model in future with replacing current models of core and pool with 3D models, and, eventually, coupling with neutronic codes for more accurate results

International Training Program in Support of Safety analysis, 3D S.UN.COP-Scaling, Uncertainty and 3D Thermal-Coupled Codes Seminars

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the ‘user effect’ and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as Barcelona (January-February 2006) and in Buenos Aires, Argentina (October 2006), being this last one requested by ARN (Autoridad Regulatoria Nuclear), NA-SA (Nucleoelectrica Argentina S.A) and CNEA (Comisión Nacional de Energía Atómica). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed ‘permanent’ stepwise approach to user training. The 3D S.UN.COP 2006 in Barcelona was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A seventh seminar is currently holding (January- February 2007) at the Texas A&M University involving more than 30 scientists between lecturers and code developers. follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Six seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering o