Fast Reactor Knowledge Preservation Efforts Overview · Fast Reactor Knowledge Preservation Efforts...

December 9-11, 2015 IAEA FR Knowledge Preservation Meeting 1

Fast Reactor Knowledge Preservation Efforts Overview

Christopher Grandy Argonne National Laboratory

IAEA Fast Reactor Knowledge Preservation Technical Exchange Meeting December 9-11, 2015 Vienna, Austria


Outline

• DOE Advanced Reactor Technologies Fast Reactor (ART-FR) Program Overview

• Fast Reactor Information and Database – FFTF Lessons Learned – R. Omberg (PNNL) – EBR-II Test Database – T. Sofu (ANL) – FFTF Reactor Test Data – R. Omberg (PNNL) – Transient Reactor Test Facility (TREAT) Database – A. Wright (ANL) – Fuels and Materials Irradiation Data – L. Yacout (ANL) – will not cover – Reliability Database – D. Grabaskas (ANL)

• Use of Fast Reactor Database – IAEA EBR-II Benchmark – T. Sofu

• Work with DOE-OSTI • Summary

ART-AFR Objective(s) and Needs

• ART-AFR Objective - Develop advanced fast reactor technology solutions to allow commercial deployment by 2050 timeframe

• Supports Nuclear Energy R&D Roadmap Objectives 2 & 3

– (2) Develop improvements in the affordability of new reactors

to enable nuclear energy to help meet the Administration’s energy security and climate change goals

– (3) Develop sustainable nuclear fuel cycles • “The overall goal is to have demonstrated the technologies

necessary to allow commercial deployment of solution(s) for the sustainable management of used nuclear fuel that is safe, economic, and secure and widely acceptable to American society by 2050.”


ART-AFR Program Activities

• ART-AFR Program is divided into various AFR-related subtasks – Fast Spectrum Reactor

• Fast Reactor Concept Development • Fast Reactor Safety • Fast Reactor Advanced Materials • Fast Reactor Inspection Technology

– Cross Cutting Research Activities • Energy Conversion Technologies • Instrumentation and Control • Licensing

– Generation IV International Support (GIF) – Modeling and Simulation (NEAMS) – high performance computing


IAEA FR Knowledge Preservation Meeting

ART-AFR Program Objectives

• ART-AFR Program Objective - Develop advanced fast reactor technology solutions to allow commercial deployment by 2050 timeframe 1. Train next generation engineers and scientists by engaging them in advanced reactor concept design

and analysis and fundamental studies that support fast reactor R&D • Allows us to transfer knowledge from seasoned reactor professionals to younger staff

2. Design and develop scalable advanced technologies for reducing the cost and/or increasing the performance of fast reactor technology

• Cost reduction • Improve safety performance • Increase system reliability

3. Preserve and manage data, knowledge, and experience related to past U.S. DOE fast reactor design, operations, tests, and component technology.

4. Re-establish the U.S. infrastructure to support the testing of advanced technologies for fast reactor applications.

5. Collaborate internationally on advanced reactor R&D through bilateral or multilateral agreements • Utilize international collaborations to leverage and expand R&D investments

December 9-11, 2015 5

Fast Reactor Knowledge Preservation and Database Goals (1/3)

• Identify and protect information that is at risk of being lost or destroyed. – EBR-II – FFTF – Fermi-I – Hallam – CRBR – Component technology development information – Commercial design studies

December 9-11, 2015 6 IAEA FR Knowledge Preservation Meeting


• Recover and organize the fast reactor knowledge to support U.S. fast reactor program (examples) – EBR-II database – FFTF database – TREAT database

• Support the preservation of knowledge related to U.S. SFR

technology and data as well as professional expertise to facilitate science-based R&D – Transfer knowledge from personnel involved in fast reactor projects to

younger staff



• Provide data for validation of the advanced analysis methods and codes for analysis of transients beyond routine operation

• Support continued participation in international collaborations in fulfillment of DOE commitments – e.g., EBR-II passive safety benchmark projects

• Make information accessible to U.S. and other Fast Reactor

technology development countries – Office of Scientific and Technical Information – IAEA CRP on EBR-II Passive Testing Benchmark

December 9-11, 2015 8 IAEA FR Knowledge Preservation Meeting

FFTF Lessons Learned


FFTF Lessons Learned Approach

• Develop a Short Write-up or Report on Each Lesson Learned

• Longer Reports Can Be Developed as Needed • Each Lessons Learned Summary Discusses:

• The Problem Found • The Resolution Method Employed

• Essentially All Will Apply to Future Design or Operating Problems

• All Are Supported by More Detailed Documents

Advanced Reactor Technology - ANL Nov 16-20 2015

10

FFTF Lessons Learned – Reports for FY15

• Bowing in Reflector Assemblies • Primary System Pressure Drop Increase • Sodium Natural Circulation and Decay Heat Removal • Secondary Sodium Flow Oscillations • Cesium Release from Failed Fuel and Transport within the Reactor

Plant • Gas Entrainment and Accumulation in Sodium and NaK Systems • Sodium Spills and Fires • Acceptance and Startup Testing

• There will be a paper by Dr. Omberg (PNNL) at ICAPP on this topic.

Advanced Reactor Technology - ANL Nov 16-20 2015

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Fast Reactor Database Development


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EBR-II Test Database

• EBR-II Test Database covers all of the experiments conducted between 1984 and 1987 during a comprehensive testing program – Data collected from the ~60 Shutdown Heat Removal Tests (SHRT),

Balance of Plant (BOP) Tests, and Plant Inherent Control Tests (PICT) • Also includes the landmark IFR inherent safety demonstration tests

– Data is organized based on the five testing windows (each with unique core, plant, and data acquisition system configuration) and test categories (such as the protected or unprotected loss of flow, loss of heat sink, or reactivity perturbation tests)

– Access to up to 900 test-specific data acquisition system instruments (grouped into 60 broad categories) to plot/tabulate and a document archive

• This database has been developed and is being used to provide information for the IAEA EBR-II CRP (will be covered later)

December 9-11, 2015 IAEA FR Knowledge Preservation Meeting

FFTF Passive Safety Database - Background

• All Passive Safety Testing (PST) related data has been preserved

– ~ 100 documents related to PSTs were identified and retrieved

– All 120 plant data tapes covering the PST tests were located, retrieved, and the data was extracted, secured, and decoded

• A web-based interface for accessing the PST data similar to that for EBR-II data is being developed

– Web-based structure was created and tested and is working

– Data retrieval and display are handled using modified versions of software developed and used during FFTF operation

– Current Capability for generating data reports for PST periods

IAEA FR Knowledge Preservation Meeting December 9-11, 2015 14

FFTF Passive Safety Database Accomplishments (1/2)

Added new data from 98 plant tapes = total 257 data tapes


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Passive Safety and Related Tests Added Tests

Cycle 7 Reactivity Feedback (12) Isothermal Physics Tests at S/U (6)

Cycle 8A Steady State Reactivity Feedback (43) Initial Ascent to Power (80)

Cycle 8B Flow Transient (3) Power-to-Melt Test (12)

Cycle 8B Delayed Pony Motor Trip (3)

Cycle 8B Steady State Natural Circulation (8)

Cycle 8C Loss of Flow Without Scram (33)

Cycle 12B1 Outage Pump Start With GEMS (10)

Acceptance Test Natural Circulation (38)

Cycle 9A Reactivity Feedback (9)

FFTF Passive Safety Database (2/2)

• Currently, any use of the raw data in the database requires expert interpretation of the data validity on a case-by-case basis.

• It is expected that this recovered information will be used in a similar manner to the EBR-II database to support International Collaborations


FFTF Data Retrieval

• If you need support of recovering information from data tapes, please contact Dr. Ron Omberg at PNNL.

• [email protected]


TREAT Database Development


Transient Reactor Test (TREAT) Facility -- used in performing approx. 900 fuel safety tests

• TREAT was designed and built by Argonne to perform power transients on reactor fuels for safety evaluations

• First critical in 1959; operated by Argonne through 1994; to be operated soon by INL

• Principal experimenter organizations were Argonne and Hanford (HEDL/WHC)

• Other experimenters: GE, ORNL, LANL • Knowledge of the tests became obscure

and effectively “buried” and practically unknown to the nuclear community today

• Need for development of a useful compendium of the historical test information

IAEA FR Knowledge Preservation Meeting 19 December 9-11, 2015

900 tests spanned multi-dimensional “matrix”

MAIN PARAMETERS • Reactor programs (LWR, LMRs, space propulsion, etc.) • Fuel materials (ceramic, metallic, cermet, etc.) • Fuel pre-test irradiation (burnup level, irradiation power,

etc.) • Fuel geometries (single pellets, single pins, pin bundles,

other) • Fuel pin design features • Cladding alloys • Coolant types (water, sodium, inert gas, etc.) • Test conditions (overpower, undercooling, TOP/TUC) • Test complexity (separate-effects tests in simple capsules

…. through loop tests with maximum-feasible prototypicality)

• Test severity (slight fuel damage, cladding breach, severe fuel disruption) IAEA FR Knowledge Preservation Meeting 20

SODIUM LOOP

December 9-11, 2015

Fuel behaviors investigated (mild fuel disruption)

Pre-failure transient fuel and cladding behaviors Fuel microstructural and composition changes due to (e.g.,) : - Atom migration - Chemical interactions - Forces/energy in crystals/grains Fission gas release from fuel grains Fission product vaporization effects Fuel fragmentation Fuel-cladding mechanical interaction Fuel-cladding chemical interactions Cladding strain Fuel melting

IAEA FR Knowledge Preservation Meeting 21

December 9-11, 2015

Fuel behaviors investigated (severe fuel disruption)

Severe accident behaviors Extensive fuel melting and cladding damage Pre-failure fuel relocations Cladding failure margins, timing relative to other events, location, mode Coolant boiling and voiding dynamics Fuel-coolant interaction effects Post-failure motions and interactions among fuel, cladding, and coolant Cladding vaporization Re-distribution of disrupted core materials


December 9-11, 2015

Test information was widely scattered

• Test information was found in a variety of places – Open literature publications – Unrestricted-distribution technical reports – Applied Technology reports – Unpublished documents

• > 4200 documents have been identified and collected to date – > 600 conference proceeding and ANS transactions – > 130 journal articles – > 900 technical reports – > 560 entries in ANL technical progress reports – > 1400 ANL internal documents (notes, memos, presentation

materials, etc.) – > 630 documents not yet categorized

• Many documents provide information on two or more tests


TREXR – Database

• >4000 documents -- categorized for all tests (spreadsheet) • Test information – basic test features categorized for all tests

(spreadsheet) • Experiment summary sheets for all tests or test groups • Digital test-data files for ~73 recent tests (with automatic graphing) TARGET USERS: • Advanced fuels developers • Reactor safety community

STRUCTURE: • Web-based architecture; draws information from the two spreadsheets • Searchable by multiple test characteristics (~120 test parameters in 19

categories) • Searchable by associated computer models and codes (~100) • Documents keyed to their technical content categories • Links to experiment document PDF files


TREXR Database Progress during FY15

• > 1400 additional documents or excerpts from published progress reports were discovered and incorporated

• > 200 experiments were added (some newly identified) • Work done by students during past two years was organized

and incorporated • Preparations were made to access and extract information from

a collection of suspect-contaminated TREAT experiment documents from Argonne hot-cell archives


Remaining tasks to complete TREXR

• Expand introductory information in home page • Finish:

– Matching documents in spreadsheet with their PDF files – Characterizing documents listed in spreadsheet – Characterizing tests listed in spreadsheet – Drafting, reviewing, and editing one-page summaries – Fixing formatting “bugs” in spreadsheets

• Collect and incorporate PDF files of photomicrographs from post-test examinations

• Include links to document publishers’ web sites • Obtain direction from DOE-NE regarding TREXR user access

authorization • Establish process and funding for TREXR future management and

user interfacing


CREDO and CREDO-II


CREDO Background (1/2)

• Throughout the 1980s and 1990s, SFR reliability data was collected and stored in a database at ORNL called the Centralized Reliability Engineering Database Organization (CREDO). – SFR reliability data includes documentation of component maintenance

events, failures, time to repair the failure, root causes, … – This data then gets summarized into component failure data expressed

in failures per hour or failures per demand for various failure modes. – This information can be used to quantify the basic event probabilities or

estimate capacity factors.

• CREDO was jointly developed by the US and Japan. This joint development stopped in 1992. Japan has expanded CREDO with information from MONJU and JOYO from this 1992 version.


CREDO Background (2/2)

• In the mid to late 1990s, after the defunding of the IFR program, ORNL disbanded the CREDO database.

• In FY14, SNL and ANL were funded to recreate the CREDO database. The initial scope of this project was to re-commission the original database and bring it unto modern database standards. – Other than occasional summary reports, no documentation remains of

the original CREDO database

• Since this time, SNL and ANL have attempted to create the database (CREDO-II) from raw reactor logs that have been preserved through Knowledge Preservation efforts.


30

Objectives – Rebuild sodium reactor component reliability database using primary sources – Support future risk assessments – Risk-inform future SFR designs

Data Sources

– EBR-II (Idaho, 1964-1994)* – Fermi Unit 1 (Michigan, 1966-1972) – FFTF (Washington, 1980-1992)* – SRE, Sodium Reactor Experiment (California, 1957-1964) – HNPF, Hallam Nuclear Power Facility (Nebraska, 1963-1964) – SEFOR, Southwest Experimental Fast Oxide Reactor (Arkansas, 1969-1972)

* Reactors included in original CREDO database.

SFR Reliability Assessments - CREDO-II

Engineering Data

Operating Data

Event Data

CREDO-II Database

Component & System

Failure Rates


CREDO Current Status

• We are currently recovering operating reports from U.S. fast reactors to: – Review these reports for equipment reliability information – Populate this database

• We are attempting to locate the old CREDO data

• Database development will take a few years to recreate if we

can gain access to suitable reports


Example of the Use of Fast Reactor Recovered Information Efforts


Benchmark Analyses of the Shutdown Heat Removal Tests Performed in the EBR-II Reactor:

IAEA Coordinated Research Project


EBR-II Reactor and Power Plant

• 20 MWe metal-fueled, sodium-cooled, pool configuration fast reactor, 1964-94

• Designed and operated by Argonne (located at former Argonne-West Idaho site)

• Complete power plant


• Impact of EBR-II operation and experiments evident in current sodium-cooled, metal fuel, pool-type designs: – GEH PRISM design (EBR-II is considered by GEH as the PRISM

prototype) – TerraPower Traveling Wave Reactor – KAERI Prototype Gen-IV Sodium-Cooled Fast Reactor project

Project Background and Goals

• Principal driver: improved evaluations of sodium-cooled fast reactor (SFR) passive safety capabilities

• Approach: validate computer simulation tools and models against data from severe accident whole-plant experimental simulations

• Data: Argonne Shutdown Heat Removal testing (SHRT) program with Experimental Breeder Reactor-II (EBR-II) in the 1980’s

• Goals: – Validate fast reactor safety analysis software – Improve design and safety analysis capabilities – Extend the validation/verification base of SFR safety analysis codes:

• LOF transients primary coolant system behavior • Fission power history during unprotected transients • Natural convective cooling and reactivity feedback mechanisms


IAEA CRP Structure

• 19 participants, representing 11 countries • Argonne is the lead technical institution, as well as a

participant, using the SAS4A/SASSYS-1 systems analysis code – Developed benchmark specifications for both transients – Provide technical support to participants – Organize and conduct Research Coordination Meetings

• Four-year program (June 2012-June 2016) • Phase 1: blind simulation of SHRT-17 (protected LOF) and

SHRT-45R (ULOF) – completed February 2014 • Phase 2: model refinement, extended comparisons against

plant data, code-to-code comparisons, uncertainty assessment – final simulation results December 2015


CRP Participants


China India Russian Federation

France Japan Switzerland Germany Republic of

Korea USA

Italy Netherlands

EBR-II CRP: Participating Organizations China: CIAE Japan: Kyushu University China: North China Electric Power Univ. Japan: University of Fukui China: Xi’an Jiatong University Korea, Republic of: KAERI France: IRSN Korea, Republic of: KINS Germany: KIT Netherlands: NRG Italy: ENEA Russian Federation: IBRAE Italy: Politecnico di Torino Switzerland: PSI Italy: UNIPI (GRNSPG) USA: Argonne National Laboratory India: IGCAR USA: TerraPower Japan: JAEA

EBR-II Plant Primary System

• 62.5 MWt • 20 MWe • Primary Sodium: 485 kg/s • Inter. Sodium: 315 kg/s • Sec. Steam: 32 kg/s • High-pressure inlet

plenum: 85% of total flow, fed inner core and extended core

• Low-pressure inlet plenum fed blanket and reflector


Benchmark Model of EBR-II Primary Vessel Components


Phase 2 Modeling Refinements

• Multidimensional effects modeling – cold pool, upper plenum, Z-Pipe, IHX primary side diffuser region, instrumented subassemblies XX09 (core location) and XX10 (reflector location)

• Cold pool modeling refinements: – 3-level 1-D model – 3-D models with coarse radial and azimuthal meshes, paired

with finer axial meshes – Computational fluid dynamics (CFD)

• Upper plenum: – Limited modeling of the baffle plate – Coarse mesh 3-D – CFD


Phase 2 Modeling Refinements (cont.)

• Primary pumps – Adjusted locked rotor flowrate threshold and locked rotor loss

coefficient – Modeled each primary pump separately

• Parametric studies on mass flow rate and pump coastdown (tuning pump parameters) – determined that both pumps locked for SHRT-17, neither locked for SHRT-45R

• Heat loss from the Z-Pipe to the cold pool • Radial heat transfer between inner and outer core • Refined core nodalization, subchannel modeling or CFD

modeling for XX09, XX10, neighboring subassemblies


Simulation Results: SHRT-17 Pump 2 Flow

• Phase 1 : many simulation results gave good agreement with the recorded data curve shape but over-predicted the magnitude by up to 70%.

• Phase 2: significant improvement for most simulations, several match the data very well from ~200 seconds on.


Phase 1

Phase 2

Simulation Results: SHRT-45R Z-Pipe Inlet Temperature

• Phase 1: All simulations agreed with the data curve shape, most overpredicted the data, no consistent agreement among simulations.

• Phase 2: Most simulations showed improved agreement with the data after 300 s.; some showed poorer agreement before 300 s.


Phase 1

Phase 2

Simulation Results: SHRT-45R IHX Inlet Temperature


• Phase 1: Results diverge more than for Z-pipe inlet; all simulations over-predict the temperatures, probably due to inadequate modeling of heat loss from Z-pipe to cold pool.

• Phase 2: Improved

results, most still over-predict recorded data.

Phase 1

Phase 2

Remaining Phase 2 Work

• Improvement in simulation results for IHX primary side inlet and outlet temperatures

• Uncertainty assessment planned using the Fast Fourier Transform Based Method (FFTBM) – Geometrical fidelity – Steady-state confirmation – Application of the FFTBM to quantify simulation accuracy

• Fourth and final Research Coordination Meeting to be held April 2016 in Vienna

• IAEA TecDoc will report the full set of CRP results – early 2017


EBR-II CRP Summary

• IAEA Coordinated Research Project on benchmark analyses of two landmark EBR-II passive safety tests is in its final year.

• Simulation results to date for phase 2 were presented and discussed at the 3rd RCM – Significant improvement over phase 1 when compared to

the recorded data – Results highlighted some areas where all simulations need

refinement – Simulation improvements continue, will be completed

December 2015 • Final RCM April 2016, TecDoc to be completed early 2017


DOE Office of Scientific and Technical Information (OSTI)


What Is OSTI?

• Free, unlimited public access to unclassified • Restricted access to classified and sensitive • Roles/responsibilities defined in DOE O 241.1B,

Scientific and Technical Information Management

“The Secretary, through the Office of Scientific and Technical Information, shall maintain within the Department publicly available collections of scientific and technical information resulting from research, development, demonstration, and commercial applications activities supported by the Department.”

OSTI’s role is anchored in law: Energy Policy Act of 2005

OSTI is the DOE office responsible for ensuring access to DOE (and predecessor) R&D results—since 1947.


What Does OSTI Do?

We make DOE R&D results findable and accessible, not just within DOE, but globally.

We make other people’s R&D results findable, bringing worldwide R&D to DOE and beyond.

Accelerating the spread of knowledge inspires everything we do at OSTI.

“If I have seen further, it is only by standing on the shoulders of giants.”

—Isaac Newton, 1676 Premise: Science advances only if knowledge is shared. Corollary: Accelerating the sharing of scientific knowledge accelerates the advancement of science.

MISSION To advance science and sustain technological creativity by making R&D findings available and useful to Department of Energy (DOE) researchers and the public.


DOE Collection • More than 300,000 electronic full-text technical reports • Over 5 million scientific e-prints • Over 2 million publicly available

citations • More than 24,000 patents • More than 500 websites & databases • Conference papers & proceedings • 1 million non-digitized documents


http://www.osti.gov/sciencecinema/

Integrates key DOE databases Covers a range of R&D results (reports, patents, citations, project summaries, eprints, etc.).

Integrates >70 nations Provides over 400 million pages of science information from databases and portals worldwide.

Integrates 13 U.S. science agencies Databases and websites offer over 200 million pages of science information.

What is special about OSTI?

• We integrate or aggregate multiple government R&D-related databases into single-search portals.

• Innovative technology drills down to selected databases and websites in parallel, then presents ranked search results.


Some International Activities

• International Nuclear Information System – INIS (www.IAEA.org/INIS) – Multilateral agreement under International Atomic Energy Agency since

1969. – World’s largest database on peaceful uses of nuclear energy and

technology. – OSTI serves as U.S. representative and input center.

• Energy Technology Data Exchange – ETDE (www.ETDE.org)

– Multilateral agreement under International Energy Agency since 1987. – Database of 4.5 million records in energy research, technology, and

development. – OSTI serves as Operating Agent.


OSTI is U.S. Focus for INIS (1/2)

Search by Common Topics = 49,840 total • Country/Org. • United States (> 7800) • Japan (> 6000) • International Atomic Energy

Agency (IAEA) (> 3500) • France (> 1900) • Germany (> 1600) • Russian Federation (> 1200)

Search by “fast reactors” = 57,323 • Country/Org. • United States (> 7300) • Japan (> 5800) • International Atomic Energy

Agency (IAEA) (> 4400) • France (> 1900) • Germany (> 1500) • Russian Federation (> 1300)


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OSTI is U.S. Focus for INIS (2/2)

“Fast Reactors” search – full text = 11,989 • Country/Org. • International Atomic Energy

Agency (IAEA) (3517) • United States (3039) • France (1116) • Japan (819) • USSR (423) • India (267)


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Summary

• ART-AFR Program is involved in a number of knowledge preservation activities

• Recovery of Information from EBR-II, FFTF, TREAT, and now other U.S. liquid metal reactors

• Recovery of Information from Office of Scientific and Technical Information (OSTI) and conversion to electronic format

• Organizing the data into electronic databases – EBR-II Plant Testing Data, FFTF Plant Testing Data, TREAT Test Data, SFR Fuels and Materials Irradiation Data, etc.

• Information is being used to support existing U.S. SFR programs along with international programs such as the IAEA CRP EBR-II Safety Benchmark


Backup


EBR-II Primary System


Benchmark Model of EBR-II Primary Vessel Components


Phase 1 Simulation Results Example

• Simulation results and recorded flow data for pump #2 during SHRT-45R • Agreement with the recorded data for several of the simulations is

within 10%


CRP Phase 2

• Argonne distributed experimental data for both transients to all participants in February 2014

• Simulation refinements by participants now underway to address modeling shortfalls identified during Phase 1: – Heat transfer modeling along the Z-pipe – Sodium mixing and heat transfer within the upper plenum – Temperature predictions throughout the IHX – Options for modeling primary pump locked rotor resistances – Cold pool stratification modeling

• Planning simulation of other SHRT transients • Third Research Coordination Meeting to be held end of

March 2015


TREAT


TREAT experiments relational (TREXR) database

• TREAT was designed for transient testing of nuclear fuels and materials under off-normal and accident conditions

• Used in performing >800 experiments from 1959 to 1994,

– most performed by Argonne

• Tests supported SFR and LWR development

• On non-operational standby since 1994, but being considered for re-use beginning in ~2019

TREAT - located at INL/MFC (formerly ANL-W, which operated it for ~35 yrs)


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December 9-11, 2015

TREAT air-cooled design allows for easy core access to a variety of in-core

experiment hardware


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December 9-11, 2015

Information from past TREAT experiments is valuable for future reactor development /licensing

TREAT Experiments ranged from investigating individual phenomena to demonstrating integral effects of multiple interactive phenomena

• Revealed fuel phenomena, interactions, thresholds, rates, etc. about which little was known or previously observed.

• Provided the basis of much of the present knowledge and understanding of transient fuel behavior and the basis for much model/code development and validation.

• Investigated transient fuel behavior phenomena investigated which are of continuing importance.

• Used techniques and approaches which can guide future experiments.

IAEA FR Knowledge Preservation

Meeting 64

December 9-11, 2015

TREAT - TREXR – Database

Target users: Nuclear fuel development and reactor safety community • Information on essentially all of the ~800 TREAT tests performed 1959-1994 • Compendium of essentially all associated journal articles, ANS

transactions, technical reports, and excerpts from progress reports (greater than 2000 documents)

• Instrument reduced data from many tests • Web-based architecture • Bibliographic information for each document • Links to document PDF files • One-page summaries for each test or small group of tests • Searchable according to:

– ~130 test parameters in 18 categories – 100 computer models and codes applied to the tests

• Under development since 2010 and will be completed in 2015.


Experiment-spreadsheet contents

• Test name, date performed, transient number • Identification number(s) of fuel pins in test (as necessary to identify

the test) • Test objectives • Fuel composition, form, core height, burnup, configuration • Cladding type • Fuel-cladding interface • Coolant / environment • Test vehicle type • Transient type • Transient measurements • Posttest analysis types • Posttest examination and measurement types, and location performed


Document-spreadsheet contents

• Filename of document PDF file

• Names of tests described in the document

• Keys indicating the types of technical content in the document

• Document bibliographical category (journal article, conf. proceeding,

report, etc.)

• Document metadata (authors, title, date, organization, report number,

conference title, volume number, page numbers, etc.)

• Names of computer models and codes(s) the document applys to the

test(s)

• OSTI reference number

• Document access designation and associated TREXR access control

code


Documents were categorized by technical content


AP Applications of test results GP General planning

CD Code description (applied to test) HO Hodoscope

DA Data from test OS Operations specifications

DC Design/construction of test rig PA Pre- & posttest analyses

DI Diagnostics to acquire test data PE Posttest examinations

ES Experiment summary evaluation SF Safety

FM Fuel motion TF TREAT facility FP Fission-product transport TH Thermal-hydraulics

FU Fuel description (pre-test)

TECHNICAL CONTENT CODES

December 9-11, 2015

Digital data from >73 tests (348 reactor runs) (mostly ANL severe-accident tests; autoplotting;

downloadable)

Test Series

No. of tests

No. of reactor runs

Year(s) performed

ANCAL 1 25 1991-92 APT TBD 33 TBD AX1 1 12 1979 C 2 2 TBD CO 7 17 1980-83 E 5 9 1968-74 EOS 3 4 1977 F 4 5 1974-80 FCT 5 5 1979 H 4 5 1972-77


Test Series

No. of tests

No. of reactor runs

Year(s) performed

HC 5 8 1976-78 HN 1 1 1977 J 1 2 1979 L 6 36 1973-78 LO 7 20 1980-83 M 8 127 1985-1993 R 3 4 1976-77 RX 2 5 1983-83 SC 4 7 1977-79 STEP

4 21 1983-85 December 9-11, 2015

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FFTF Passive Safety Database

• A similar effort (to the EBR-II database) is underway at PNNL to develop a database for FFTF passive safety demonstration tests – Objectives are to verify natural circulation as

a reliable means to safely remove decay heat, extend passive safety experience to a large-size LMR, develop and test passive safety enhancements

– Will be useful to support future simulation and modeling efforts: • Potential benchmarks of advanced simulation models against realistic

operational data • Well-characterized environment for verifying coupled thermal

hydraulic/neutronic/mechanical codes


EBR-II Fuels Irradiation and Physics Analysis Database Development

(FIPD)


Introduction to FIDB

• Originally created to support the neutronics simulation of the Integral Fast Reactor (IFR) fuel cycle.

• Contains the data required to simulate the UZr/UPuZr fuel flow into EBR-II, the fuel irradiation during the reactor operation, and the fuel flow out of EBR-II. – Database covers EBR-II operation from 08/03/1984 to 09/30/1994 – EBR-II operation data was organized into different runs from Run 130A to

Run 170B. – Depletion calculations were performed for every EBR-II run with REBUS

and RCT. – Auxiliary codes were developed to facilitate data processing. – PADB consists of physically measured data, code calculated data, and

auxiliary codes.

72 December 9-11, 2015 IAEA FR Knowledge Preservation Meeting

EBR-II Fuels Irradiation & Physics Database (FIPD)

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Motivation: – Knowledge base for U-Zr metallic alloy fuel

• PIE reports, drawings, experiments qualification report, reactor information … etc.

– EBR-II experiments: Prototype fuel behavior, fuel swelling and

restructuring, lead metal fuel experiment … etc.


EBR-II Fuels Irradiation & Physics Database

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Objective: – Create a web-accessible relational database as an archive of information from EBR-

II fuels irradiation experiments as well as calculations based information in an appropriate form:

• For use by fuels and reactors analysts/ modelers • Directly useful for models / codes validations activities

Database End-User

– FIPD is important to industry and institutions with interest in developing metallic fuel based fast reactors:

• Terrapower, Toshiba, GE, ARC, KAERI …etc

– FIPD supports validation and calibration of current fuel performance codes (e.g. LIFE-METAL) which is important for the industry and relevant institutions.


EBR-II Fuels Irradiation & Physics Database (FIPD)

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I Experiment

General info

Fabrication data

Physics data

TH data Documents

Pin data

PIE Fuel/ clad properties SE-2 /

CFD

LIFE-METAL

- ART/ NEAMS/ modelers/ industry

- Fuel qualification - Safety evaluation - Continuum models - Transient codes - NRC licensing


Data in (FIPD) (I)

• Run level data: – Starting and ending time of a run. – Total energy generated during a run. – Control and safety rod position at the

beginning and end of a run. – Core loading pattern.

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Covered by FIDB (97 runs)


Data in (FIPD) (II)

• Subassembly level data: – Structural masses of axial regions of

different subassembly types – Number of fuel pins – Notch position for indication of subassembly

orientation, and only 2 orientations are possible.

– Fuel composition of irradiated subassembly (calculated from REBUS).

• Fuel element level data: – Fuel pin position relative to the notch. – Fuel composition for UZr/UPuZr fuel: fresh,

reconstituted, and irradiated (calculated from RCT).

– fuel pin designs including slug outer diameter, cladding inner and outer diameter, fuel height, cladding material, wire wrap mass.

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SS304: 634.32 g

304SS: 880.34 g INC : 815.42 g NA : 59.35 g VOID : 151.37 cc

304SS: 6912.70 g NA : 278.20 g DU : 2521.64 cc

304SS: 272.54 g

304SS: 2984.85 g

Outer Blanket UDEPUNF5 December 9-11, 2015 IAEA FR Knowledge Preservation Meeting

EBR-II FIDB Database: Current Status & Development

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Database structure has been developed and is currently being populated

About 16 experimental assemblies have been included in the database – Data generation of EBR-II experiments: X419, X420, X421, X423, X425, X429, X430, X441,

X447, X448, X449, X451, X489, X492, X496, and X501

Generating data of additional EBR-II experiments based on anticipated needs of ART program or to support code validation activities

The selected experiments represent remaining key steady state metal fuel experiments

conducted at EBR-II – X431 and X432 blanket fuel with HT9 cladding and high smeared density – X452, X453, X454, X455 are U-10Z fuel experiments with impurities to simulate impact of

casting impurities on FCCI, FCMI, and fuel behavior.


Fast Reactor Knowledge Preservation Efforts Overview · Fast Reactor Knowledge Preservation Efforts...

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