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Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV)

    The Expert Group will deal with the activities associated with the certification of experimental data and benchmark models. The objectives are to provide with:
  • access to certified experimental data from the contributions of individual member countries
  • guidance and recommendations for developing benchmark models from certified experimental datasets
  • access to standardized benchmark models with detailed uncertainty evaluations and uncertainty methodology guidelines
  • recommendations and guidelines for the range of applicability of the certified experimental datasets
  • guidelines and consensus recommendations for validating multi-physics simulations

Working Group on Fuel Safety (WGFS)

    The objectives are to promote a better understanding on the handling of leaking fuel in power reactors, discuss and review the current practices in member countries and to summarise results and draw conclusions to help in decisions on the specification of reactor operation conditions with leaking fuel rods and on the handling of leaking fuel after removal from reactor. Planned activities for the WGFS include:
  • RIA fuel rod codes benchmark (recent experiments performed in the NSRR and CABRI test reactors)
  • Mechanical testing of fuel cladding for RIA applications
  • Reviewing design-basis accidents, especially LOCA and RIA
  • Periodically assessing and reviewing fuel safety criteria
  • Investigating advanced fuel designs and effect on safety margins

Expert Group on Uncertainty Analysis in Modelling (UAM-LWR) Coupled Multi-physics and Multi-scale LWR analysis

    Uncertainty analysis in modelling (UAM) is to be further developed and validated on scientific grounds in support of its performance, in addition to LWR best-estimate calculations for design and safety analysis. There is a need for efficient and powerful analysis methods suitable for such complex coupled multi-physics and multi-scale simulations. The proposed sequence of benchmarks will address this need by combining the expertise in reactor physics, thermal-hydraulics etc. and uncertainty and sensitivity analysis, and will contribute to the introduction of advanced/optimised uncertainty methods in best-estimate reactor simulations.

State-of-the-art report on simulation capabilities of 3D system-scale thermal-hydraulic codes (3DSYSTH)

    Most of the recent SYStem-scale Thermal-Hydraulic codes (3DSYSTH) propose limited prediction capabilities for dealing with 3D TH phenomena. These codes are used for 3D plant analysis without a full understanding of the limitations of their 3D capabilities. Hence, this activity aims to determine the state-of-the-art simulation capability of SYS TH codes in relation to 3D phenomena relevant to nuclear safety. The product will be a state-of-the-art report aiming to:
  • provide an overview of existing 3D TH safety issues in NPPs and of 3D simulation capabilities covering models and correlations in systems codes
  • identify code limitations for NPP applications
  • review the existing PIRTs and experimental database relevant to 3D TH safety issues
  • identify the validation database for 3D SYS TH code

Analysis of Options and Experimental Examination of Fuels for Water-Cooled Reactors with Increased Accident Tolerance (ACTOF)

    Globally, there is a great deal of experience with the performance of reactor fuel in off-normal conditions. Theoretical studies and experiments have been performed and there have been excursions form normal operating conditions in a few power reactors. During such an excursion, the difference between an incident of limited or no consequence and a severe accident, such the one at Fukushima, depends on the conditions in the reactor and the performance of the fuel under those conditions. This Project will explore the potential to design and operate advanced fuel types that are intended to be more tolerant of severe accident conditions whilst retaining the capability of current fuel designs for safe operation under normal operation and anticipated transient conditions.

NEA Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) Project

    The ATLAS project is aimed at topics of high safety relevance for both existing and future nuclear power plants. It is important because the use of computer codes is required in safety evaluation of LWRs in order to simulate plant behaviour during DBAs and design extension conditions (DECs). This involves complex multi-dimensional single-phase and two-phase flow conditions. The objective of ATLAS is to generate system-integral and separate-effect experimental database to validate predictive capability and accuracy of computer codes and models for complex phenomena of high safety relevance to thermal-hydraulic transients in DBA and DEC.

Task Group on a Phenomena Identification and Ranking Table for Spent Fuel Pools in Loss-of-Cooling and Loss-of-Coolant Accident Conditions

    The objective is to conduct a (PIRT) on Spent Fuel Pool (SFP) in LOCA conditions based on expert opinion elicitations. A particular emphasis will be made on mitigation strategies. The objectives of the PIRT would be to:
  • Systematically identify phenomena that are both of high safety importance and of high uncertainty
  • Determine if the efficiency of potential mitigation means, such as water spray, are enough assessed
  • The PIRT would therefore cover:
  • the behaviour of the SFP prior to fuel uncovery
  • the thermal-hydraulic behaviour of the SFP
  • the fuel behaviour during fuel uncovery
  • SFP recriticality and accident mitigation
  • the influence of air ingress on the fuel degradation
  • the fission-product releases and transport

Oskarshamn-2 (O2) BWR Stability Benchmark for Coupled Code Calculations and Uncertainty Analysis in Modelling

    This benchmark defines a coupled code problem for further validation of thermal-hydraulics system codes for application to BWR stability based on actual plant data from the Oskarshamn-2 NPP which experienced a stability event on 1999. A loss of feed-water pre-heaters and control system logic failure resulted in a condition with a high feed-water flow and low feed-water temperature without reactor scram. In addition to the initiating event, an interaction of the automatic power and flow control system caused the plant to move into low flow – high power regime. Combination of above events culminated in diverging power oscillations which triggered automatic scram at high power. Five linear stability measurements before the event and five linear stability measurements after the event are included in the benchmark.

NEA Primary Coolant Loop Test Facility (PKL-3) Project

    The PKL-2 test programme is investigating safety issues relevant for current PWR plants as well as for new PWR design concepts and will focus on complex heat transfer mechanisms in the steam generators and boron precipitation processes under postulated accident situations. Tests addresses current safety issues related to BDBA transients with significant core heat-up, i.e. station blackout scenarios or loss of coolant accidents in connection with failure of safety systems. Without any adequate accident management (AM) procedures the postulated courses of events would lead to a severe accident scenario with core damage. Both scenarios will be connected with an assessment of the performance of the Core Exit Temperature (CET) which is used as criterion for the initiation of AM-measures involving emergency operating procedures and/or severe accident management measures.

Benchmark Analyses of an EBR-II Shutdown Heat Removal Test

    The IAEA Coordinated Research Projects addresses Shutdown Heat Removal Tests (SHRT) performed at the Experimental fast Breeder Reactor EBR-II within the framework of the US Integral Fast Reactor development and demonstration programme. The CRP will improve the participants' simulation capabilities in the various fields of research and design of sodium cooled fast reactors through data and codes validation and qualification. The scope of this CRP is twofold: firstly, validation of the state-of-art liquid metal cooled fast reactor codes and data used in neutronics, thermal hydraulics and safety analyses, and, secondly, training of the next generation of fast reactor analysts through international benchmark exercises.

Status report on informing Severe Accident Management guidance and actions through analytical simulation

    The result of analytical simulation of Severe-Accident Management (SAM) actions may help identify gaps or potential weaknesses in the existing SAMG and thus help improve or refine it. The status report will present the best and recommended practices regarding the use of analytical simulations as one of the means to validate SAM. The main objectives are:
  • To provide a basis for consistent definitions of concepts of “verification” and “validation” of SAM actions
  • Describe several existing practices aiming at ensuring the correctness, usability and efficiency of SAM
  • With respect to analytical simulations, the report will provide an overview of national examples of past and ongoing assessments of SAM through modelling of operator actions by integrated codes.

PREMIUM: A Benchmark on the Quantification of the Uncertainty of the Physical Models in the System Thermal-Hydraulic Codes - Methodologies and Data Review

    The PREMIUM benchmark is a WGAMA activity that follows up the BEMUSE programme dedicated to quantification of the uncertainties of the physical models in thermal-hydraulic system codes. BEMUSE clearly showed that quantification of input-parameter uncertainty is a key point in any uncertainty analysis since these input uncertainties are propagated through the considered system code in order to obtain the uncertainty of the output parameters. PREMIUM is specifically devoted to this issue. To this end, PREMIUM considers concrete cases related to the physical models related to prediction of core reflood based on FEBA and PERICLES experiments.