|
|
|
NEA-1765 IRPHE2009-HANDBOOK. (Abstract last modified 21-MAR-2009)
1.
NAME OR DESIGNATION - IRPHE2009-HANDBOOK. 2.
COMPUTER FOR WHICH THIS PACKAGE IS DESIGNED -
To request or retrieve programs click on the one of the active versions below.
A password and special authorization is required. Explanation of the status codes.
Machines used:
Package-ID Orig.Computer Test Computer
NEA-1765/05 Many Computers Many Computers
3.
DESCRIPTION - 8.
RELATED INFORMATION - 9.
STATUS 10.
REFERENCES - 12.
PROGRAMMING LANGUAGE -NEA-1765/05: 14.
OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS - 15.
NAME AND ESTABLISHMENT OF AUTHORS - 16.
MATERIAL AVAILABLE -NEA-1765/05: 17.
CATEGORIES - Keywords: evaluated data, experiment, fast reactors, high-temp reactor, pressurized water reactor, reactor physics
Program-name Package-ID Status
IRPHE2006-HANDBOOK NEA-1765/01 Obsolete
IRPHE2007-HANDBOOK NEA-1765/02 Obsolete
IRPHE2007-ARCHIVES NEA-1765/03 Obsolete
IRPHE2008-HANDBOOK NEA-1765/04 Obsolete
IRPHE2009-HANDBOOK NEA-1765/05 Tested
The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data.
********************************************************************************
The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/index.shtml
********************************************************************************
The evaluation process entails the following steps:
1. Identify a comprehensive set of reactor physics experimental measurements data,
2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility,
3. Compile the data into a standardized format,
4. Perform calculations of each experiment with standard reactor physics codes where it would add information,
5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data.
The International Handbook of Evaluated Reactor Physics Benchmark Experiments 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 personal to validate calculational techniques.
The 2009 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 36 different experimental series that were performed at 21 different reactor facilities. 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 current edition contains the following:
FUND - Fundamental:
------------------
- ATR-FUND-RESR-001 - CRIT (new)
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 (MIX-MISC-FAST-001)
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 - Gas Cooled (Thermal) Reactor:
----------------------------------
- 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 (new)
Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Fully-Loaded Core)
HWR - Heavy Water Moderated Reactor:
-----------------------------------
- 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 (new)
Evaluation of Measurements Performed on the French High Flux Reactor (HFR)
LMFR - Liquid Metal Fast Reactor:
--------------------------------
- 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 (new)
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 (new)
SNEAK 7A and 7B: Pu-Fueled Fast Critical Assemblies in the Karlsruhe Fast Critical Facility (new)
- ZPPR-LMFR-RESR-001 - CRIT-REAC-RRATE (new)
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 (new)
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 (new)
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 (new)
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 - Light Water Moderated Reactor:
-----------------------------------
- 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 (new)
Isothermal Experiment of the IPEN/MB-01 Reactor
- KRITZ-LWR-RESR-001 - CRIT-BUCK-RRATE (new)
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 (new)
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 (new)
KRITZ-2:13 Experiment on Regular H2O/Fuel Pin Lattices with Low Enriched Uranium Fuel at Temperatures 22.1 ?C and 243 ?C
PWR - Pressurized Water Reactor:
-------------------------------
- CREOLE-PWR-EXP-001 - CRIT-COEF-RRATE-MISC
CREOLE PWR Reactivity Temperature Coefficient Experiment - UOX and MOX up to 300C in EOLE
- VENUS-PWR-EXP-001 - RRATE-POWDIS (new)
VENUS-1 PWR UO2 Core 2-Dimensional Benchmark Experiment
- VENUS-PWR-EXP-003 - RRATE-POWDIS (new)
VENUS-3 PWR UO2 Core 3-Dimensional Benchmark Experiment
- VENUS-PWR-EXP-005 - CRIT-SPEC-POWDIS (new)
Experimental Study of the VENUS-PRP Configuration No. 9
VVER - VVER Reactor:
-------------------
- PFACILITY-VVER-EXP-001 - CRIT-RRATE
VVER Physics Experiments: Hexagonal (1.27-vm 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
International Criticality Safety Benchmark Evaluation Project (ICSBEP).
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-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
NEA-1765/01: 25-MAY-2007 Obsolete
NEA-1765/02: 04-MAR-2008 Obsolete restricted
NEA-1765/03: 07-MAR-2008 Obsolete
NEA-1765/04: 14-APR-2009 Obsolete
NEA-1765/05: 21-MAR-2009 Tested restricted
NEA-1765/05:
- International Handbook of Evaluated Reactor Physics Benchmark Experiments
NEA/NSC/DOC(2006)1, March 2009 Edition (OECD/NEA)
This project is being carried out under the auspices of the OECD Nuclear Energy Agency
Contact at NEA:
Computer Program Service
OECD/NEA Data Bank
12 boulevard des Iles
92130 Issy les Moulineaux (France)
Contact at INL (USA):
J. Blair Briggs
Idaho National Laboratory
2525 North Fremont
P.O. Box 1625
Idaho Falls, ID 83415-3890 (United States)
The International Handbook of Evaluated Reactor Physics Benchmark Experiments 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 of America.
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.
DVD with reactor physics evaluations
- Y. Integral Experiments Data, Databases, Benchmarks