HWR Related Activities for 2011 TWG-HWR Meeting Related Activities for 2011 TWG-HWR Meeting Jong-Ho CHOI Division of Nuclear Power IAEA 1 International Atomic Energy Agency List of

HWR Related Activities for 2011 TWG-HWR Meeting

Jong-Ho CHOIDivision of Nuclear Power

IAEA

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International Atomic Energy Agency

List of HWR Related Activities1. ICSP on Comparison of HWR Thermalhydraulic Code

Predictions with SBLOCA Experimental Data2. CRP on Benchmarking Severe Accident Computer

Codes for HWR Applications3. Workshop on Good Practices in HWR Operation4. The Role HWR In the Sustainable Utilization of

Fissionable Resources5. Workshop on the Prediction of Axial and Radial Creep in

HWR Pressure Tubes (TM-41824, 16-18 Nov. 2011)6. TM on Fuel Design and Licensing of Mixed Cores for

Water Cooled Reactors (TM-41825, 12-14 Dec. 2011)7. Planned New Activities for 2012-13

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ICSP on Comparison of HWR Thermalhydraulic Code Predictions

with SBLOCA Experimental Data (1)

• ICSP on LBLOCA analysis with RD-14M was conducted by TWG-HWR (1999-2003, TECDOC-1395)

• Outline for new ICSP on SBLOCA was presented and endorsed by TWG-HWR in 2007 June meeting. TWG recommended to conduct ICSP for two cases with different break sizes.

• Objectives of ICSP:– Improve understanding of important phenomena expected to occur in HWR

SBLOCA transients– Evaluate code capabilities to predict these important phenomena, their practicality

and efficiency, by simulating an integral experiment– Suggest necessary code improvements or new experiments to reduce

uncertainties– Improve the capability of code users by sharing the experience

• Coordinated jointly with Safety Assessment Section of the Division of Nuclear Installation Safety

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ICSP on Comparison of HWR Thermalhydraulic Code

Predictions with SBLOCA Experimental Data (2)

• Participants and Computer Codes

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Institute Country Computer CodeCNEA Argentina CATHENA Mod-3.5c/Rev 0AECL Canada CATHENA Mod-3.5d/Rev 2Tsinghua U. China CATHENA Mod-3.5d/Rev 2AERB India RELAP5/Mod3.2NPCIL India ATMIKAKAERI Rep. of Korea CATHENA Mod-3.5d/Rev 2KINS Rep. of Korea MARS-KSCNPP Romania CATHENA Mod-3.5d/Rev 1




• ICSP Scope– Selection and distribution of experiment case

• RD-14M Tests B9006: 7-mm Inlet Header SBLOCA with ECI• RD-14M Test B9802: 3-mm Inlet Header SBLOCA

– Set up analysis methodology– Steady state simulation– Blind simulation– Comparison of blind simulation results– Release of experimental data– Open simulation– Comparison and Identification of Deviation– Documentation

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• Overall Schedule of ICSP– 1st Meeting: Vienna, Austria, 2007 November

• Lessons learned from previous ICSP on LBLOCA• Important phenomena for HWR SBLOCA (subset of 52 comparison

parameters among 600 measurements)• Scope and schedule

– Distribution of facility description, test summay, initial and bounday conditions: early 2008

– 2nd Meeting: Winnipeg, Canada, 2008 August• Review/discussion of steady state calculation results• RD-14M test facility visit• Ground rules and common assumptions for blind calculation

– 3rd Meeting: Vienna, Austria, 2009 August• Review/discussion of blind calculation results

– 4th Meeting: Daejeon, Korea, 2010 November• Review/discussion of final blind and open calculation results• Formulation of conclusions and recommendations

International Atomic Energy AgencyResults for B9006 7

Top FES




• Lessons Learned from ICSP– ICSP offered a constructive platform giving an unique opportunity to code

developers/experts, users and experimentalists to jointly verify and validate their thermal-hydraulic codes.

– The comparison between code results and test data is simplified and improved if instrument location is given due consideration in the nodalization of the entire facility.

– User effects• ECC system modelling (valve logic, flow resistances, ECC tank model

and/or pressure boundary condition)• PHTS pump characteristics • Break discharge modelling• Grouping of FES within channels to allow temperature stratification. Those

that did not allow stratification (e.g. by lumping all heated FES into one group) predicted maximum FES temperatures closer to those measured.

• Familiarity with code and facility

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• Lessons Learned from ICSP– Code deficiencies and need for capability improvement

• All codes were able to replicate the following challenging phenomena during simulation of SBLOCA scenarios:

• Two-phase pressure drop• Low flows and flow reversal in parallel channels (the quality of agreement

is strongly dependent on matching steady-state flow distribution)• Loop refill as a result of ECC injection • Timing of LPECC start• Channel voiding• Condensation in boilers (steam generators)• FES dryout

• All codes experienced difficulties in the following areas (B9802):• Precise matching of break discharge flow rate (even with tuning of

discharge models)• Dryout/rewet cycles, prior to PDO heat transfer, in the channels• Slow heat-up in PDO regime (FES temperatures in PDO operation were

generally overpredicted)9






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• Lessons Learned from ICSP– Experimental considerations and ICSP specification

• Exact physical description and operation of ECC system and secondary side

• PHTS pump characteristics• Factors affecting differential pressure measurements (in particular those

spanning a significant elevation difference)• Unambiguous units for pressure measurements and pressure-related trip

set-points and logic (MPa(a) or MPa(g)) • Calibration, accuracy limits, and uncertainty of void meters and flow

meters• Timing (including instrument delays) of power and pump rundowns and

other trips, such as valve open and close times

• Current ICSP Status– Draft document is ready for final edition with IAEA format




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• Summary and Conclusions– All calculations were capable of achieving steady state conditions consistent with

the experimental data apart from deviations in flow distribution among the individual parallel channels in each pass.

– For SBLOCA, channel voiding is not a concern as much as for LBLOCA.– Improvements in PDO heat transfer and in the critical break discharge models are

required.– ECC effectiveness for channel refilling and SG effectiveness for condensation are

important phenomena that affect FES heatup. – All main phenomena (e.g. break discharge, coolant voiding, pressure drop, boiling

and condensation heat transfer, and temperature excursion in the heated sections) are qualitatively captured by the participants.

– The application of codes developed outside the HWR technology did not show any special deficiency in the comparison with the present experimental database. However, the existence of parallel channels, and the potential of flow reversal in some of them, is untypical in other PWR reactor systems and requires special attention in the modeling.

– The performed activity is relevant in assessing the capabilities of codes and permitted the quantification of the amount of discrepancy between measured and calculated values.


CRP on Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications (1)

• Specific Research Objectives:– to improve safety for currently operating HWRs and to facilitate more

economic and safe designs for future plants • Participating Organizations and Computer Codes

– AECL (Canada): MAAP-CANDU– Shanghai Jiao Tong Univ. (SJTU, China): RELAP5, SCDAP– BARC1 (India): RELAP5, MELCOOL– BARC2 (India): RELAP5, SCDAP, ASTEC– NPCIL (India): ATMIKA.T, CONTACT, SEVAX– KAERI (Rep. of Korea): ISSAC– Politehnica Univ. of Bucharest (PUB, Romania): RELAP5, SCDAP, COUPLE

• Expected Outcomes– Improved understanding on HWR severe accident phenomena– Consensus on HWR severe accident scenario– Advanced information on computer code capabilities– Recommendations for improvements and subsequent research

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• CRP Activities:– Assessment of the existing models, correlations, experiments, and computer

codes– CANDU 6 benchmark analysis for station blackout

• Establish criteria for major sequential failures: fuel failure, fuel channel failure, fuel channel disassembly, core collapse, calandria vessel failure and containment failure, and reactor vault failure.

• Phase 1 : Accident initiation to fuel channel dryout• Phase 2 : Fuel channel dryout to core collapse • Phase 3 : Core collapse to calandria vessel failure• Phase 4 : Calandria vessel failure to containment failure

– New experiment and Benchmark analysis for experiment: MFMI tests (AECL), TROI tests (KAERI), Channel Heatup and Debris Bed Heat Transfer tests (BARC)

– Documentation

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• Parameters to be compared for Phase 1:– PHTS pressure and temperature (outlet header)– PHTS mass flow rate– Maximum sheath temperature– Mass inventories in the primary HTS (including pressurizer) and steam generator– The time when channel void reaches 0.9 for the entire length of channel for the first time in a chan

nel– PHTS water level (pressurizer)– Time when: liquid relief valve first opens, MSSV opens, first fuel channel rupture, and moderator r

upture disk rupture.– Time when first channel uncovery occurs– Secondary side pressure transient (SG)– Amount of hydrogen produced– Accumulated heat transfer to the moderator form the fuel channels (MJ) – Amount of fuel failure and fission product– Moderator level, pressure and temperature– Heat transfer through SG (MW)

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• Parameters to be compared for Phase 2:– Moderator level in the calandria vessel– Maximum temperature of any fuel channel and maximum sheath temperatu

re– Amount of hydrogen produced – Time of core collapse– Steam flow rate from the calandria vessel to the containment– The amount of heat transferred to the moderator from the fuel channels (kW

)– The amount of heat transferred to the reactor vault water (kW)– Calandria vessel wall temperature

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• Parameters to be compared for Phase 3:– Reactor vault water level– Time of the calandria vessel water depletion– Maximum calandria vessel wall temperature– Maximum debris temperature– Total heat load on the calandria vessel (kW)– Molten fraction of the debris– Time and location of calandria vessel failure– Mass of relocated core material – Amount of hydrogen produced– Amount of fission products released

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• Parameters to be compared for Phase 4:– Amount of fission products released at the containment boundary– Containment pressure– Amount of hydrogen produced– Amount of concrete ablation– Amount of non-condensable and steam produced– Temperature of the concrete (containment wall)– Temperature of the debris bed (calandria vault)– Failure time of containment– Mass of steam condensing on the containment wall– Time when calandria vault water inventory is depleted

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• Some preliminary results

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• RCM:– The 1st RCM: IAEA-HQ, Vienna, Austria, February 2009– The 2nd RCM: NPCIL, Mumbai, India, May 2010– The 3rd RCM: KAERI, Daejeon, Rep. of Korea, September 2011

• Current CRP Status:– Most participants completed the preliminary analysis up to phase 4– Submitted preliminary analysis report

• Future plan– Perform final analysis– Comparison of results– Preparation of draft TECDOC

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Workshop on Good Practices in HWR Operation (1)

• Endorsed by TWG-HWR in 2005 December meeting

– Positive experience from operating HWRs should be documented in an integrated manner and made available to the HWR community

• Workshop was proposed to implement this activity by TWG-HWR in 2007 June meeting

• AECL hosted the 1st Workshop on “Best Practices in HWR Operation” in Toronto, 16-19 September 2008

– TECDOC-1650, Good Practices in HWR Operation, June 2010, Summary and papers presented at 1st workshop

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• The 2nd Workshop– Hosted by KHNP, 12-14 April 2011, Gyeongju, Rep. of Korea– Attended by 19 international and 23 local participants– 21 good practices were shared

• 36 month outage cycle (Darlington, OPG, Canada)• lowering the excess reactivity to 1.5 mk (TQNPC, China)• MODAR (MOdel Detection And Repositioning) development for detection and

repositioning of TFGS (Bruce Power and AECL, Canada)• Recent planned outage of Wolsong Unit 2 with the period of 25.36 days (from 12

November – 7 December 2010) earlier than the plan of 26.3 days (KHNP, Korea)

– http://www.iaea.org/NuclearPower/Technology/Meetings/2011-April-12-14-WS.html

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The 2nd WorkshopGyeongju, Korea12-14 April 2011

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The Role HWR In the Sustainable Utilization of Fissionable Resources

• Objectives: – To identify design, engineering and technological advances that can enhance the

role of HWR in improving the utilization of fissionable resource;– Specifically, to identify possible fuel cycles synergisms among various reactor

designs that can help achieve an optimum utilization of global fissionable resource, and

– To identify challenges to the materialization of these advances, as well as possible solutions and R&D needs to overcome them

• First Consultancy Meeting: 11-13 Dec 2006 at Mumbai, India• Second Consultancy Meeting: 4-7 Dec 2007 at Beijing, China• No real progress until July 2009• TWG in July 2009 recommended to consider inclusion in the new

activity, Use of Thorium Fuel in Existing and Future HWRs.• After 2009 TWG meeting further discussion between NPTDS/NENP

and NFCMS/NEFW led to the completion of this task first by lead of NPTDS (Mr. Tyobeka) with strong support of NFCMS (Mr. Basak)

• Draft report outline was prepared

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Workshop on the Prediction of Axial and Radial Creep in HWR Pressure Tubes

• TM-41824, Workshop on the Prediction of Axial and Radial Creep in HWR Pressure Tubes, 16-18 November 2011, Room A0478, IAEA-HQ, Vienna, Austria

• Objectives of the Meeting– to exchange the information on dimensional change of pressure tubes and its

impact on the plant operation and safety– to understand Member States’ R&D status on pressure tube creep and

interest/intention for a new IAEA CRP to setup a data base and to develop a prediction model for HWR pressure tube creep

– to set up a road map to prepare a CRP proposal • Expected Technical Topics:

– Measurement of pressure tube creep– Research on pressure tube creep phenomena – Improvements in pressure tube manufacturing, and the effect on pressure tube

deformation– Effect of pressure tube creep on plant operation and safety– Licensing issues related to pressure tube creep– Development of prediction model for pressure tube creep– Operating concerns to minimize pressure tube creep– Operational strategy to compensate pressure tube creep

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TM on Fuel Design and Licensing of Mixed Cores for Water Cooled Reactors (1)

• TM-41825, FUEL DESIGN AND LICENSING OF MIXED CORES FOR WATER COOLED REACTORS, 12-14 December 2011, IAEA Headquarters, Vienna, Austria

• TM to evaluate the merits and challenges of increasing the Uranium enrichment above 5% was replaced by this TM:

– To reflect the change of interest after Fukushima accident– To facilitate the establishment of IAEA nuclear fuel bank– TWG-FPT suggest the change first

• Objectives– Information exchange on experience in design, licensing and operation of mixed

cores – Detailed review of safety and licensing issues of mixed cores – Examination of new approaches or analytical tools for modelling mixed cores to

perform core physics, structural and thermal hydraulic analyses

• Organized with cooperation of NSNI and NEFW

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TM on Fuel Design and Licensing of Mixed Cores for Water Cooled Reactors (2)

• Topics to be covered:– Experience in obtaining improved fuel cycle economics by operating with mixed

cores– Safety and licensing aspects of mixed cores – Fuel design requirements for mixed cores– Analytical tools for modelling mixed cores to calculate their core physics, core

thermal-hydraulics and structural behaviour– Experience of operation with lead test assemblies– Fuel management with mixed cores– Experience in obtaining improved fuel performance with mixed cores with new fuel

assembly designs– Problems that have been encountered with mixed cores related to fuel assembly

bowing and/or fuel cladding failures

• Note Verbale is in the internal clearance process and will be sent to 29 countries with operating NPP plus Iran, UAE, Turkey, Vietnam, Jordan, Lithuania, Italy

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Planned New Activities for 2012-13 P&B

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Coordinated Research Projects (CRPs)

• Prediction of Diametral Creep of HWR Pressure Tube (HWR-4, 2010)– Will be decided based on TM-41824 (Workshop on the Prediction

of Axial and Radial Creep in HWR Pressure Tubes, 16-18 November 2011) considering:

• The number of countries/institutes interested in this CRP

• Probability to achieve valuable outputs through CRP

• Benchmarking CFD codes used in the design of advanced water cooled reactors

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International Collaborative Assessments (ICAs)/Technical Meetings/Workshops/Courses (1)

• The 3rd Workshop on Good Practices in HWR Operation, scheduled in 2013 (HWR-5, 2010)– Need to find host organization

• Facilitate technology advances for design and deployment of new water cooled reactors– TM to share best practices and experiences in the design and

validation of passive safety systems, and in the optimization of the use of active and passive safety systems

– TM to share best practices and experiences in the design and validation of severe accident mitigation systems

– TM to share best practices and experiences in the design of seismic protection systems

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International Collaborative Assessments (ICAs)/Technical Meetings/Workshops/Courses (2)

• Facilitate the development of advanced modelling and simulation tools for the design and deployment of new water cooled reactors– Workshop on advanced code suite for design, safety analysis and

operation of HWRs (HWR-3, 2010)• Performance, safety and fuel cycle options of the world’s HWRs,

Inter-regional TC project, merged into “Supporting Member States in the Evaluation of Nuclear Reactor Technology for Near-Term Deployment”, 2012-2015 (HWR-6, 2010)

• Facilitate technology development medical isotope production in commercial NPPs– TM on production of medical isotopes in HWRs (HWR-9, 2010)

• Course on science and technology of supercritical water cooled reactors

• Course on natural circulation phenomena and passive safety systems in advanced water cooled reactors

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Documents

HWR Related Activities for 2011 TWG-HWR Meeting Related Activities for 2011 TWG-HWR Meeting Jong-Ho CHOI Division of Nuclear Power IAEA 1 International Atomic Energy Agency List of