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International Reactor Physics Benchmark Experiments (IRPhE) Project

The International Reactor Physics Benchmark Experiments (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows:

  • maintaining an inventory of the experiments that have been carried out and documented;
  • archiving the primary documents and data released in computer-readable form;
  • promoting the use of the format and methods developed and seek to have them adopted as a standard. 
For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in:
  • compiling experiments into a standard international agreed format;
  • verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility;
  • analysing and interpreting the experiments with current state-of-the-art methods;
  • publishing electronically the benchmark evaluations.
The expert group will: 
  • identify gaps in data and provide guidance on priorities for future experiments; 
  • involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods;
  • provide a tool for improved exploitation of completed experiments for Generation IV reactors;
  • co-ordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; 
  • keep a close link with the working parties on scientific issues of reactor systems  (WPRS), , the expert group on reactor-based plutonium disposition (TFRPD), now integrated into WPRS, scientific issues of the fuel cycle (WPFC) and the working party on international evaluation co-operation (WPEC).
The Secretariat of the group is provided by the OECD/NEA Data Bank, who is in charge of the management of the material released to the project. The Technical Review is chaired and co-ordinated by J.B. Briggs from INL, USA.

IRPHE meeting for the March 2010 Handbook edition Paris, France the 20-23 Ocotober 2009

Registration (only for expert group members)

Current activities

The group is currently:

  • maintaining an international inventory of relevant reactor physics experiments and measurements, preserving in computer readable form the primary documentation and data describing the facilities, characterising the experimental techniques, the experimental results and interpretation thereof;
  • developing and maintaining an international database containing a sub-set of high priority reactor physics benchmark specifications derived from the experiments performed at various nuclear facilities around the world, relevant for or including data from power reactors (e.g. Advanced Gas Reactors, Fast Reactors, Generation-IV Reactors, Water Reactors, etc.)
The following type of measurements are included:
  • fundamental mode lattice experiments;
  • heterogeneous core configurations;
  • power reactor start-up data;
  • core follow experiments; 
  • specific applications experiments (e.g. fission product integral data, irradiation experiments).
The benchmark specifications and experimental data are intended for use by nuclear reactor physicists and engineers to validate current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques.

Fourth edition and availability


The fourth edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments was published in March 2009 (ISBN 978-92-64-99055-5) and contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personnel to validate calculational techniques. This Edition contains data from 36 different experimental series that were performed at 21 different reactor facilities. A few of these evaluations are published as DRAFT documents only. Draft documents have been reviewed by the IRPhEP Technical Review Group (TRG); however, all action items could not be completed or reviewed in time for the final publication or, in most cases, the TRG felt it necessary to review the revised evaluations before giving final approval. The Handbook is organized in a manner that allows easy inclusion of additional evaluations, as they become available. Additional evaluations are in progress and will be added to the handbook periodically.

The Handbook is published in electronic format (pdf files) on DVD, where the experiments are grouped into evaluations, categorised by (1) Reactor Name, (2) Reactor Type, (3) Facility Type, and (4) Measurement Type. 

The Handbook was prepared by a working party comprised of experienced reactor physics personnel from Belgium, Brazil, Canada, P.R. of China, France, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Slovenia, Switzerland, United Kingdom, and the United States.
 
The Handbook can be requested by accessing the corresponding IRPhEP web pages filling in the REQUEST FORM. It is distributed on DVD. The IRPhEP Handbook is available to authorised requesters from the OECD member countries and to contributing establishments from non-OECD countries. Other requests are handled on a case by case basis.  Restrictions, Disclaimer, Feedback

Handbook contents March 2009 Edition


Evaluations

TITLE

 

 

FUND

 
  ATR-FUND-RESR-001
  CRIT
 Advanced Test Reactor:  Serpentine Arrangement of Highly Enriched Water-Moderated Uranium-Aluminide Fuel Plates Reflected by Beryllium
  BFS1-FUND-EXP-001
  CRIT-SPEC-COEF-RRATE		 Experimental Program Performed at the BFS-97, -99,  -101 Assemblies - Critical Experiments with Heterogeneous Compositions of Plutonium, Depleted Uranium Dioxide and Polyethylene
  BFS1-FUND-EXP-002
  CRIT-SPEC-REAC-RRATE		 Experimental Program Performed at the BFS-42 Assembly - k-infinity Experiments for 238U in Fast Neutron Spectra:  Measurements with Plutonium Mixed with Depleted Uranium Dioxide and Polyethylene
  BFS1-FUND-EXP-003
  CRIT-SPEC-COEF-RRATE		 Experimental Program Performed at the BFS-57and -59 Assemblies - Critical Experiments with Heterogeneous Compositions of Enriched Uranium or Plutonium, Depleted Uranium Dioxide and Polyethylene
  BFS2-FUND-EXP-001
  CRIT-SPEC-REAC		 Experimental Program Performed at the BFS-31 Assemblies - k-infinity Experiments for 238U in Fast Neutron Spectra:  Measurements with Plutonium Mixed with Depleted Uranium Dioxide

 

 

GCR

 
  ASTRA-GCR-EXP-001
  CRIT		 Graphite Annular Core Assemblies with Fuel Elements Containing UO2 Coated Fuel Particles
  HTR10-GCR-RESR-001
  CRIT		 Evaluation of the Initial Critical Configuration of the HTR-10 Pebble-Bed Reactor
  HTTR-GCR-RESR-001
  CRIT-SUB-REAC-COEF-KIN-RRATE
 Evaluation of the Start-Up Core Physics Tests at Japan’s High Temperature Engineering Test Reactor (Fully-Loaded Core)

 

 

HWR

 
  DCA-HWR-RESR-001
  CRIT-SPEC-RRATE		 Deuterium Critical Assembly with 1.2% Enriched Uranium Varying Coolant Void Fraction and Lattice Pitch
  HFR-HWR-RESR-001
  CRIT
 Evaluation of Measurements Performed on the  French High Flux Reactor (HFR)

 

 

LMFR

 
  BFS1-LMFR-EXP-001
  CRIT-SPEC-COEF-KIN-RRATE		 BFS-73-1 Assembly:  Experimental Model of Sodium-Cooled Fast Reactor with Core of Metal Uranium Fuel of 18.5% Enrichment and Depleted Uranium Dioxide Blanket
  BFS2-LMFR-EXP-002
  CRIT-SPEC-RRATE		 BFS-62-3A Experiment:  Fast Reactor Core with U and U-Pu fuel of 17% Enrichment and Partial Stainless Steel Reflector
  JOYO-LMFR-RESR-001
  CRIT-REAC-COEF		 Japan's  Experimental Fast Reactor JOYO MK-Icore:  Sodium-Cooled Uranium=Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO2 Blanket
  ZEBRA-LMFR-EXP-001
  CRIT-SPEC-REAC-RRATE		 Fast Critical Experiments in Plate and Pin Geometry Form. The ZEBRA CADENZA Cores, Assemblies 22, 23, 24 and 25
  ZEBRA-LMFR-EXP-002
  CRIT-SPEC-REAC-RRATE		 The ZEBRA MOZART Programme Part 1. MZA and MZB, ZEBRA Assemblies 11 and 12
  ZEBRA-LMFR-EXP-003
  CRIT-REAC-RRATE		 The ZEBRA MOZART Programme Part 2. MZC and the Control Rod Studies ZEBRA Assembly 12
  ZPR-LMFR-EXP-001
  CRIT-SPEC-REAC-RRATE		 ZPR-6 Assembly 7:  A Cylindrical Assembly with Mixed (Pu,U)-Oxide Fuel and Sodium with a Thick Depleted-Uranium Reflector
  ZPR-LMFR-EXP-002
  CRIT-REAC-RRATE
 ZPR-6 Assembly 7 with High Pu-240:  A Fast Reactor Core With Mixed (Pu,U)-Oxide Fuel and Sodium and High Pu-240 Zone
  SNEAK-LMFR-EXP-001
  CRIT-SPEC-REAC-RRATE
 SNEAK 7A and 7B:  Pu-Fueled Fast Critical Assemblies in the Karlsruhe Fast Critical Facility
  ZPPR-LMFR-RESR-001
  CRIT-REAC-RRATE		 ZPPR-10A Experiment:  A 650 MWe-Class Sodium-Cooled MOX-Fueled FBR Homogeneous Core Mock-Up Critical Experiment with Two Enrichment Zones and Nineteen Control Rod Positions
  ZPPR-LMFR-EXP-002
  CRIT-SPEC-REAC-RRATE
 ZPPR-9 Experiment:  A 650 MWe-Class Sodium-Cooled Mox-Fueled FBR Core Mock-Up Critical Experiment with Clean Core of Two Homogeneous Zones
  ZPPR-LMFR-EXP-003
  CRIT-SPEC-REAC-RRATE-MISC
 ZPPR-18A Experiment:  A 1,000 MWe-Class Sodium-Cooled MOX-Fueled FBR Core Mock-Up Critical Experiment with Two-Homogeneous Zones and Control-Rod Withdrawal, where Enriched Uranium is used with the Shape of a Sector in the Outer Core
  ZPPR-LMFR-EXP-004
  CRIT-SPEC-REAC-RRATE
 ZPPR-19B Experiment:  A 1,000 MWe-Class Sodium-Cooled MOX-Fueled FBR Core Mock-Up Critical Experiment with Two-Homogeneous Zones and Control-Rod Withdrawal, where Plutonium and Enriched Uranium are used Mixing in the Outer Core

 

 

LWR

 
  CROCUS-LWR-RESR-001
  CRIT-REAC-KIN		 Kinetic Parameters and Reactivity Effect Experiments in CROCUS
  DIMPLE-LWR-RESR-001            
  CRIT-BUCK-SPEC-REAC-COEF-RRATE    
Light Water Moderated and Reflected Low Enriched Uranium (3 wt.% 235U) Dioxide Rod Lattices DIMPLE S01
  DIMPLE-LWR-RESR-002
  CRIT-BUCK-SPEC-COEF-RRATE		 Light Water Moderated and Reflected Low Enriched Uranium (3 wt.% 235U) Dioxide Rod Lattices DIMPLE S06
  IPEN/MB01-LWR-RESR-001
  CRIT-COEF-KIN-RRATE		 Isothermal Experiment of the IPEN/MB-01 Reactor
  KRITZ-LWR-RESR-001
  CRIT-BUCK-RRATE
 KRITZ-2:19 Experiment on Regular H2O/Fuel Pin Lattices with Mixed Oxide Fuel at Temperatures 21.1 and 235.9 °C
  KRITZ-LWR-RESR-002
  CRIT-BUCK-RRATE
 KRITZ-2:1 Experiment on Regular H2O/Fuel Pin Lattices with Low Enriched Uranium Fuel at Temperatures 19.7 °C and 248.5 °C
  KRITZ-LWR-RESR-003
  CRIT-BUCK-RRATE
 KRITZ-2:13 Experiment on Regular H2O/Fuel Pin Lattices with Low Enriched Uranium Fuel at Temperatures 22.1 °C and 243 °C

 

 

PWR

 
  CREOLE-PWR-EXP-001
  CRIT-COEF-RRATE-MISC		 CREOLE PWR Reactivity Temperature Coefficient Experiment - UOX and MOX up to 300 °C in EOLE
  VENUS-PWR-EXP-001
  RRATE-POWDIS
 VENUS-1 PWR UO2 Core 2-Dimensional Benchmark Experiment
  VENUS-PWR-EXP-003
  RRATE-POWDIS
 VENUS-3 PWR UO2 Core 3-Dimensional Benchmark Experiment
  VENUS-PWR-EXP-005
  CRIT-SPEC-POWDIS
 Experimental Study of the VENUS-PRP Configuration No. 9

 

 

VVER

 
  PFACILITY-VVER-EXP-001
  CRIT-RRATE		 VVER Physics Experiments: Hexagonal (1.27-cm Pitch) Lattices of U(4.4 wt.% 235U)O2 Fuel Rods In Light Water, Perturbed by Boron, Hafnium, or Dysprosium Absorber Rods, or by Water Gap With/Without Aluminium Tubes
  ZR6-VVER-EXP-001
  CRIT-BUCK-SPEC-REAC-COEF-RRATE  
The VVER Experiments: Regular and Perturbed Hexagonal Lattices of Low-Enriched UO2 Fuel Rods in Light Water


 New or modified in the March 2009 edition


Additional evaluations planned for inclusion into the Handbook March 2010 Edition


  • TCA-LWR-EXP-001 CRIT-COEF Temperature Effect on Reactivity of Light Water Moderated UO2 Core with Soluble Poisons
  • IPEN/MB01-LWR-RESR-001 Reactor Physics Experiments in the IPEN/MB-01 Research Reactor Facility, Spectral Indices
  • VENUS-PWR-EXP-002 RRATE-POWDIS    VENUS-2 PWR MOX Core Measurements
  • VENUS-PWR-EXP-004 RRATE-POWDIS    Experimental Study of the VENUS Configuration No. 7
  • VENUS-PWR-EXP-005 RRATE-POWDIS    Experimental Study of the VENUS Configuration No. 17
  • VHTRC-GCR-RESR-001 Temperature Effect on Reactivity of VHTRC-1 Core
  • HTR-PROTEUS-GCR-EXP-001 CRIT-REAC-KIN-RRATE-MISC, HTR-PROTEUS pebble bed reactor experiments Shut-down Rod Worths in LEU-HTR Configurations
  • HTTR-GCR-RESR-002 Start-up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)
  • B&WSSCR-PWR-EXP-001 B&W-SSCR, Spectral Shift Reactor Lattice Experiments, Configuration I
  • ZEBRA-RESR-EXP-001 K-Infinity Experiments In Fast/Intermediate Neutron Spectra for Various Fissile Materials - ZEBRA Core 8
  • Fast Flux Test Facility (FFTF) 400 MW, Na-cooled, MOX-fueled fast reactor initial isothermal fully-loaded-core critical
  • ZPPR-13A:  650 MWe class, radially-heterogeneous clean core in JUPITER: very weak core-coupling
  • ZPPR-17A: 650 MWe class, axially-heterogeneous clean core in JUPITER: medium characteristics between homogeneous and radially-heterogeneous core from core-coupling
  • JOYO - Burn-up reactivity measured in 50MWth power ascension test, 50MWth 1st and 2nd cycles, and 75MWth 1st-4th  cycles
  • INL Power Burst Facility – IMF-Type ternary ceramic fuel.  Criticality and kinetics data


Recently added primary documentation archives (December 2008)

  • IRPHE/B&W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments
  • IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan
  • IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database
  • IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility
  • IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation
  • IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility
  • IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation
  • IRPHE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents
  • IRPHE-HTR-ARCH-01, Archive of HTR Primary Documents
  • IRPHE/AVR, AVR High Temperature Reactor Experience, Archival Documentation
  • IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters
  • IRPhE/BERENICE, effective delayed neutron fraction measurements
  • IRPhE-TAPIRO-ARCHIVE, fast neutron source reactor primary documents, reactor physics experiments

Other benchmarks in reactor physics (coupled neutronics/thermal-hydraulics - coupling core-plant)

  • ZZ 3DLWRCT, 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks
  • ZZ BFBT, OECD/NEA-US/NRC NUPEC BWR Full-size Fine-mesh Bundle Tests Benchmark
  • ZZ BWRSB-FORSMARKS, Stability Benchmark Data from BWR FORSMARKS 1 and 2
  • ZZ BWRSB-RINGHALS1, Stability Benchmark Data from BWR RINGHALS-1
  • ZZ BWRTT, BWR Turbine Trip Transient Benchmark Based on Peach-Bottom 2
  • ZZ PBMR-400: PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark
  • ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics
  • ZZ V1000CT-1 VVER-1000 Main Coolant Pump (MCP) Switching On Benchmark
  • ZZ WPPR Pu Recycling Benchmark Results
  • ZZ UAM-LWR Uncertainty Analysis in Modelling, Coupled Multi-physics and Multi-scale LWR analysis

Evaluation Guide - Revision 8.9 (20 January 2006)
Format Template


Co-operation 

  • IRPhE co-operates and co-ordinates its work with the ICSBEP: International Criticality Safety Benchmark Evaluation Project.

Acknowledgments

Much of the work so far realised by the International Reactor Physics Experiments Evaluation Project (IRPhEP), in particular, the evaluation and review of selected benchmark experiments, was possible thanks to substantial funding provided by the Government of Japan. Other countries, currently Belgium, Brazil, Canada, P.R. of China, France, Germany, Hungary, Japan, the Republic of Korea, the Russian Federation, Sweden, Switzerland, the United Kingdom, and the United States of America have contributed evaluations, reviews and data at their own expense. Overall technical coordination of the IRPhEP is directly supported by the United States Department of Energy’s Office of Nuclear Energy with significant in-kind contributions from the parallel OECD NEA International Criticality Safety Benchmark Evaluation Project (ICSBEP), supported in the United States by the Department of Energy’s Office of Facility Management and ES&H Support.

The Fifth IRPhE Technical Review Meeting was held in Paris on 20-24 October 2008. During this meeting, experiments evaluations were reviewed that will be added in the third edition of the IRPhEP Handbook. The next such meeting will be held in Paris from 20-23 October 2009.

Related links

Password-protected area for IRPhE evaluators and reviewers

Archived NEACRP and NSC documents


Last updated: 31 March 2009