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.

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.

Seventh edition and availability

The seventh edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments was published in May 2012 (ISBN 978-92-64-99168-2) 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 56 different experimental series that were performed at 32 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 2012 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
NRAD-FUND-RESR-001 CRIT	Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with uranium(20)-Erbium-Zirconium-Hydride Fuel
PBF-FUND-RESR-001 CRIT	Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water
RHF-FUND-RESR-001 CRIT-REAC	Evaluation of Measurements Performed on the French High Flux Reactor (RHF)
SCCA-FUND-EXP-001 CRIT-SPEC-REAC-RRATE	Critical Configuration and Physics Measurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-cm Pitch)
ZEBRA-FUND-RESR-001 CRIT-SPEC	K-Infinity Measurements in Zebra Core 8
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)
HTTR-GCR-RESR-002 CRIT-REAC-RRATE	Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)
HTTR-GCR-RESR-003 CRIT-REAC-RRATE	Evaluation of Zero-Power, Elevated-Temperature Measurements at Japan's High Temperature Engineering Test Reactor
PROTEUS-GCR-EXP-001 CRIT	HTR-Proteus Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio
VHTRC-GCR-EXP-001 CRIT- COEF - Draft	Temperature Effect on Reactivity in VHTRC-1 Core
HWR
DCA-HWR-RESR-001 CRIT-SPEC-RRATE	Deuterium Critical Assembly with 1.2% Enriched Uranium Varying Coolant Void Fraction and Lattice Pitch
ZED2-HWR-EXP-001 CRIT	28-Element Natural UO2 Fuel Assemblies in ZED-2
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
BFS1-LMFR-EXP-002 CRIT-SPEC-REAC-KIN-RRATE	BFS-61 Assemblies: Experimental Model of Lead-Cooled Fast Reactor with Core of Metal Plutonium-Depleted Uranium Fuel and Different Reflectors
BFS2-LMFR-EXP-001 CRIT-SPEC-RRATE	BFS-62-3A Experiment: Fast Reactor Core with U and U-Pu fuel of 17% Enrichment and Partial Stainless Steel Reflector
FFTF-LMFR-RESR-001 CRIT-SPEC-REAC-COEF-MISC	Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor
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
SNEAK-LFMR-EXP-001 CRIT-BUCK-SPEC-COEF-KIN-RRATE-MISC	SNEAK 7A and 7B PU-Fueled Fast Critical Assemblies in the Karlsruhe Fast Critical Facility
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 12/4 and 12/5
ZPPR-LMFR-EXP-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
ZPPR-LMFR-EXP-005 CRIT-SPEC-REAC-RRATE	ZPPR-10B Experiment: A 650 MWe-Class Sodium-Cooled MOX-Fueled FBR Homogeneous Core Mock-Up Critical Experiment with Two Enrichment Zones, Seven Control Rods and Twelve Control Rod Positions
ZPPR-LMFR-EXP-006 CRIT-SPEC-REAC-RRATE	ZPPR-10C Experiment: A 800 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-007 CRIT-SPEC-REAC-RRATE-MISC	ZPPR-13A Experiment: A 650 MWe-Class Sodium-Cooled MOX-Fueled FBR Radial Heterogeneous Core Mock-Up Critical Experiment with Central Blanket Zone and Two Internal Blanket Rings
ZPPR-LMFR-EXP-008 CRIT-SPEC-RRATE	ZPPR-18C Experiment: A 1,000 MWe-Class Sodium-Cooled MOX-Fueled FBR Homogeneous Core Mock-Up Critical Experiment in the State of Removal of One of Eighteen Half-Inserted Control Rods, where Enriched Uranium is used with the Shape of a Sector in the Outer Core
ZPPR-LMFR-EXP-009 CRIT-SPEC-REAC-RRATE-MISC	ZPPR-17A Experiment: A 650 MWe-Class Sodium-Cooled Mox-Fueled FBR Axial Heterogeneous Core Mock-Up Critical Experiment with Central Internal Blanket Zone
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
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- 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
TCA-LWR-EXP-001 COEF	Temperature Effects on Reactivity in Light Water Moderated UO2 Core with Soluble Poisons at TCA
PWR
CREOLE-PWR-EXP-001 CRIT-COEF-RRATE-MISC	CREOLE PWR Reactivity Temperature Coefficient Experiment - UOX and MOX up to 300 °C in EOLE
SSCR- PWR-EXP-001 CRIT-BUCK-SPEC	B&W Spectral Shift Reactor Lattice Experiment: A 484 Uranium Rods Critical Experiment with Infinite Radial Reflector
VENUS-PWR-EXP-001 RRATE-POWDIS - Draft	VENUS-1 PWR UO2 Core 2-Dimensional Benchmark Experiment
VENUS-PWR-EXP-003 RRATE-POWDIS - Draft	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
RBMK(CF)
RBMKCF-RBMK-EXP-001 CRIT	RBMK Graphite Reactor: Uniform Configurations of U(1.8, 2.0, or 2.4% 235U) U(1.8, 2.0, or 2.4% 235U)O2 Fuel Assemblies, and Configurations of U(2.0% 235U)O2 Assemblies with Empty Channels, Water Columns, and Boron or Thorium Absorbers, with or without Water in Channels
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
LR0-VVER-EXP-001 CRIT - Draft	VVER Physics Experiments: Hexagonal Lattices (1.22-cm Pitch) of Low-Enriched U(2.0 - 3.3 WT.% U235)O2 Fuel Assemblies in Light Water with Central Control-Assembly Mockup
New or modified in the May 2012 edition

Additional evaluations planned for inclusion into the Handbook March 2013 Edition

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

Recently added primary documentation archives (Last update 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.