The Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS) and its Expert Groups co-ordinate benchmark activities on reactor single- and multi-physics. The benchmark activities serve the WPRS objectives:
The activities address the the following reactor types:
The WPRS organises an annual WPRS Benchmarks Workshop (typically 1 week with 2 parallel tracks) in which the benchmark teams exchange on the status of their simulations and discuss benchmark specification updates.
The table below shows all ongoing and completed benchmarks under the WPRS guidance. Please follow the links to learn more about options to acccess the data and to participate in ongoing activities.
|Benchmark Title||Reactor Type||Benchmark Type||Status|
|Benchmark on Pu Burner Fast Reactor||FR||Neutronics||Published via NEA DB|
|Benchmarks on Pu recyling||PWR; FR||Neutronics||Published via NEA DB|
|BFBT: NEA/NRC Benchmark based on NUPEC BWR Full-size Fine-mesh Bundle Tests (BFBT)||BWR||Thermalhydraulics; Coupled Neutronics/Thermal-hydraulics||Published via NEA DB|
|Benchmark on Burst Fission Gas Release||Fuel Performance||Ongoing|
|BWR-TT: Boiling Water Reactor Turbine Trip||BWR||Coupled Neutronics/Thermal-hydraulics||Published via NEA DB|
|C5G7-TD: The Deterministic Time-Dependent Neutron Transport Benchmark C5G7-TD without Spatial Homogenisation||PWR||Neutronics; Multiphysics||Ongoing|
|CTH: McMaster CANDU TH||CANDU||Thermalhydraulics||Ongoing|
|FHR: Fluoride-salt-cooled High temperature Reactor||MSR||Neutronics||Ongoing|
|Forsmark 1&2 BWR Stability Benchmark||BWR||Coupled Neutronics/Thermal-hydraulics||Published via NEA DB|
|HTGR T/H Benchmark||HTGR||Thermalhydraulics||Proposal Stage|
|International Comparison of a Depletion Calculation Benchmark on Fuel Cycle Issues||PWR||Neutronics||NEA/NSC/DOC(2013)1|
|KALININ-3 Coolant Transient Benchmark||VVER||Coupled Neutronics/Thermal-hydraulics; Neutronics; Thermalhydraulics; Multiphysics||Published via NEA DB|
|LFR Benchmark: Lead-cooled Fast Reactor Benchmark||LMFR||Neutronics||Ongoing|
|LMFR T/H: Liquid Metal Fast Reactor Core Thermal-Hydraulics Benchmark (LMFR T/H)||LMFR||Thermalhydraulics||Ongoing|
|MHTGR-350: Benchmark of the Modular High-Temperature Gas-Cooled Reactor-350 MW Core Design||HTGR||Neutronics||Completed|
|MOX/UO2 Transient: Pressurised Water Reactor MOX/UO2 Core Transient Benchmark||PWR||Neutronics||Published via NEA DB|
|MPCMIV: Multi-physics Pellet Cladding Mechanical Interaction Validation||Test reactor||Multiphysics||Ongoing|
|MSLB: Pressurised Water Reactor Main Steam-Line Break Benchmark||PWR||Coupled Neutronics/Thermal-hydraulics||Published via NEA DB|
|O2: Oskarshamn-2 (O2) BWR Stability Benchmark||BWR||Neutronics; Thermalhydraulics; Multiphysics||Ongoing|
|PBMR: Coupled Neutronics/Thermal-hydraulics Transients Benchmark - The PBMR-400 Core Design||HTGR||Coupled Neutronics/Thermal-hydraulics||Published via NEA DB|
|PCMI: Pellet Cladding Mechanical Interaction Benchmark||PWR; BWR||Fuel Performance||Ongoing|
|PSBT: International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests||PWR||Thermalhydraulics||Published via NEA DB|
|Ringhals-1&2 Stability Benchmark||BWR||Coupled Neutronics/Thermal-hydraulics||Published via NEA DB|
|ROSTOV-2: Benchmark on reactivity compensation of boron dilution by stepwise insertion of control rod cluster into the VVER-1000 core||VVER||Neutronics; Thermalhydraulics; Multiphysics||Ongoing|
|TVA Watts Bar Unit-1 Multi-Physics Benchmark||PWR||Neutronics; Thermalhydraulics; Multiphysics||Ongoing|
|UAM LWR: Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM)||PWR; BWR; VVER||Neutronics; Multiphysics; Fuel Performance; Thermalhydraulics||Ongoing|
|UAM SFR - Subgroup on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM)||LMFR||Fuel Performance; Multiphysics; Neutronics; Thermalhydraulics||Ongoing|
|V1000CT -VVER-1000 Coolant Transient benchmark||VVER||Coupled Neutronics/Thermal-hydraulics; Neutronics; Thermalhydraulics||Published via NEA DB|
Working area for benchmark participants: MyNEA (password protected)
The Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of Light Water Reactors (LWRs) is an international high-visibility benchmark for uncertainty analysis in best-estimate coupled code calculations for design, operation, and safety analysis of LWRs. The annual workshops are attended by many experts in industry, research institutes, national laboratories, academia, and government agencies.
To ensure reliable modelling of neutron physics within a state-of-the-art transient code, the neutron kinetics part of such a code should be based on the full-scale calculation of the space-time neutron kinetics equations without use of the diffusion approximation and spatial homogenisation.
Lead-cooled Fast Reactors (LFR) are rather new concepts which have gathered increasing international attention after the Generation-IV International Forum (GIF) selected them as promising candidates for a new generation of nuclear energy systems. The LRF physics benchmark is based on the Advanced LFR European Demonstrator (ALFRED) design and consists of a neutronics and thermal hydraulics stage with each three benchmark phases related to pin-cell, sub assembly/super-cell and whole-core simulations.
The Liquid Metal Fast Reactor (LFMR) is one of the next generation reactor designs. This benchmark consists of steady-state numerical predictions of Texas A&M University (TAMU) separate effect tests and of numerical predictions of the Thermal Hydraulic Out of Reactor Safety (THORS) integral effect tests and comparison to experimental results.
A number of tests with detail well documented neutronics and thermal-hydraulics measurements data have been performed at the Rostov Unit 2 (Rostov-2) nuclear power plant (NPP). The reactor type is a VVER-1000 with fuel assemblies of type TBC-2M, which enable an 18-month fuel cycle length.
The Subgroup on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM) was formed to check the use of best-estimate codes and data.