Safety Strategy of JSFR establishing In-Vessel Retention of Core Disruptive Accident

Safety Strategy of JSFR establishing In-Vessel Retention of Core Disruptive Accident Yoshiharu TOBITA Advanced Nuclear System R&D Directorate Japan Atomic Energy Agency (JAEA) FR13, aris France, March 2013

Fast Reactor Cycle Technology Development roject 1. Introduction The safety strategy of JSFR against CDA IVR (In-vessel retention) of ATWS by appropriate design measures. Against the mechanical failure of RV: Elimination of re-criticality Against the thermal failure of RV: Cooling and retention of core materials by in-vessel debris tray LOHRS type accidents should be practically elliminated by providing divers and multiple design measure to prevent CDA. 1

Fast Reactor Cycle Technology Development roject Accident scenario in ATWS Initiation of CDA The design measures: Maximum void worth less than ~ 6 $, etc Discharge path Initiating phase No energetics Expansion Energetics Energetics Transition phase Gradual progression Structure response SA can wall Grid spacer Core Inner duct Design measures: Introduction of FAIDUS Early discharge Large core pool No energetics Material relocation Heat removal Discharge path Cross section Reference for JSFR FAIDUS option 2

Fast Reactor Cycle Technology Development roject Concept of competition between fuel dispersal and FCI-void development (simplified expression to clarify dependency on design parameters) Competition Onset of fuel motion 3

Fast Reactor Cycle Technology Development roject JSFR design parameters 4 Fuel specific power is kept above 40kW/kg (lower specific heat tends to lead to higher power transient) 3 Limited core height which provides enhanced axial fuelworth gradient 2 High failure threshold with annular pellet 1 Limited sodium void worth ( < 6$ including uncertainty) ~ These characteristics are basically in common with the EFR design strategy [H. Sztark, et al., FR91] 4

Case ID Fast Reactor Cycle Technology Development roject arametric study using SAS4A for JSFR ULOF I Analytical condition Sodium void reactivity Doppler coefficient Fission gas as driving force for fuel dispersal Cladding strength REF Nominal Nominal Nominal Nominal VOID Nominal 1.2 DO Nominal Nominal 0.8 FG Nominal Nominal x 0.5 UTS * Nominal Nominal x 0.5 ALL Nominal 1.2 Nominal 0.8 Nominal x 0.5 Nominal x 0.5 * low Ultimate Tensile Strength [I. Sato, et al., Development of Severe Accident Evaluation Technology (Level 2 SA) for Sodium-cooled Fast Reactors (2) Identification of Dominant Factors in Initiating hase of Unprotected Events, ICA 09, Tokyo, May, 2009] 5

Fast Reactor Cycle Technology Development roject ρ FCI void disp 170o ρ fuel dispersal t disp An example of a mild sequence (ρ void, max =5$) 6

Fast Reactor Cycle Technology Development roject Summary of SAS4A results Case ID eak ρ net eak power Core av. temp. REF 0.9660$ 164 o 2749 K VOID 0.9903$ 459 o 2665 K DO 0.9938$ 402 o 2622 K FG 0.9670$ 185 o 2715 K UTS 0.9671$ 172 o 2748 K ALL 0.9930$ 551 o 2680 K No energetic sequence was calculated within the present parametrics [I. Sato, et al., Development of Severe Accident Evaluation Technology (Level 2 SA) for Sodium-cooled Fast Reactors (2) Identification of Dominant Factors in Initiating hase of Unprotected Events, ICA 09, Tokyo, May, 2009] 7

Fast Reactor Cycle Technology Development roject Accident scenario in ATWS Initiation of CDA The design measures: Maximum void worth less than ~ 6 $, etc Discharge path Initiating phase No energetics Expansion Structure response Energetics Energetics (re-criticality) Transition phase Gradual progression SA can wall Grid spacer Core Inner duct Design measures: Introduction of FAIDUS Early discharge Large core pool No energetics Material relocation Heat removal Discharge path Cross section Reference for JSFR FAIDUS option 8

Inner duct Fast Reactor Cycle Technology Development roject Assessments on effectiveness of the design measure Event sequence in the early-discharge phase Qualitative evaluation based on experimental data Shielding material UAB Leak tight blockage Core A A LAB Inner duct A-A SA can wall Close end Release of inter-/intragranular fission gas, mixing with molten fuel, and formation of two phase foamy state ressurization of core region Heat transfer to SA can wall and inner duct Inner-duct failure preceds SA can wall failure. Upward discharge of molten fuel driven by fission gas flow without blockage formation Remaining solid fuel Subsequent materialrelocation phase To the upper coolant plenum Shielding material UAB Core LAB Gas plenum To the lower coolant plenum 9

Fast Reactor Cycle Technology Development roject Fuel Discharge Analyses using the SIMMER-III code Objective and calculation procedure Objective Confirmation of the elimination of re-criticality by early fuel discharge in FAIDUS Calculation procedure The core is divided into several groups based on power. Fuel discharge analysis in the representative subassemblys in each group by 2-dimensional analysis using SIMMER-III. The discharge behavior in each SAs are synthesized to whole core behavior. Reactivity change is evaluated by 3-D neutronics calculation and point kinetics calculates the feadback to power transient. The above procedure is repeated several times until convergence between the discharged mass and power transient is reached. B B B B Core center 10 B B : rimary CR : Back-up CR : Radial blanket 1 2 3 4 5 6 7

Ratio of fuels remaining in each Ratio of fuels remaining region [-] in each region [-] Normalized poewr [-] Normalized power [/ 0 ] Ratio Ratio of of fuels fuels remaining in in the the core core[-] Net reactivity [$] Fast Reactor Cycle Technology Development roject Fuel Discharge Analyses using the SIMMER-III code Results The higher the thermal output was, the more fuel discharged. Negative reactivity insertion by fuel discharge reduced the reactivity and the power of the reactor. The core remnant consists of solid fuel and its low mobility prevents re-criticality. 1 0.8 0.6 0.4 0.2 0 1 2 3 40 5 6 7 Region No. 1 0.95 0.9 0.85 0.8 0.75 0.7 1000 x x 100 10 1 Region 1 Region 2 Region 3 Region 4 Region 5 Region 6 Region 7 Total No fuel discharge CW failure 0.1 18 19 20 21 22 23 24 25 Time from LOF onset [s] Time after LOF onset [s] 2 0-2 -4-6 -8-10 -12-14 Net reativity [$] 11

Fast Reactor Cycle Technology Development roject Accident scenario in ATWS Initiation of CDA The design measures: Maximum void worth less than ~ 6 $, etc Discharge path Initiating phase No energetics Expansion Structure response Energetics Energetics (re-criticality) Transition phase Gradual progression SA can wall Grid spacer Core Inner duct Design measures: Introduction of FAIDUS Early discharge Large core pool No energetics Material relocation Heat removal Discharge path Cross section Reference for JSFR FAIDUS option 12

Fast Reactor Cycle Technology Development roject IVR-Failure Factors in Material-Relocation hase Required conditions and contributing factors Required condition against mechanical boundary failure: No severe energetics caused by the motion of core-remaining materials Axial movement of solid-fuel/molten-steel remaining in the core Radial motion of re-melting fuel Chemical/thermal stability of B 4 C in disrupted core Required condition against thermal boundary failure: No thermal failure of RV caused by the contact of discharged molten materials Quenching of discharged materials due to thermal interaction with sodium RV (Reactor Vessel) Outline of material-relocation phase 13

Reactivity insertion Fast Reactor Cycle Technology Development roject Design Measure against Severe Energetics redicted event progression Early stage of Material-relocation phase Late stage of Material-relocation phase ±0$ Falling of absorber rod (B 4 C) Falling of dispersed fuel in inner core Collapses of UCS & outer core Failure of CRGTs (1) Diffusion of B/Fe eutectic around backup CRGT (2) Discharge of molten fuel through primary CRGT (3) Discharge of molten steel through primary CRGT rogression of fuel melting (due to decay power) (3) (1) (2) Time after the start of material-relocation phase End of Early-discharge phase (6 sec after the power peak in Initiating phase) 14

Reactivity [$] Fast Reactor Cycle Technology Development roject Design Measure against Severe Energetics Evaluation of reactivity transient Early stage of Material-relocation phase (About 30 sec) Late stage of Material-relocation phase (Several minutes) 0$ -5$ -10$ -15$ -20$ -25$ -30$ Falling of absorber rod (B 4 C) -5.5$ Falling of dispersed fuel in inner core -18$ Collapses of UCS & outer core Diffusion of B/Fe eutectic by backup-crgt failure -2.5$ -4.4$ -9.7$ -6.1$ Uncertainty of fuel-compaction in outer core (un-melted SA) Uncertainty of B/Fe-diffusion region Discharge of molten-fuel/steel by primary-crgt failure -15$ Uncertainty of steel amount remaining in the core region after fuel discharge About 100 sec; Backup-CRGT failure 2 3 min; Fluidization of the core-remaining fuel due to decay heat 3 4 min; rimary-crgt failure -23$ -26$ End of Early-discharge phase (6 sec after power peak in initiating phase) End of Early-discharge phase (~6 sec after power peak in Initiating hase) Time after the start of material-relocation phase 15

Fast Reactor Cycle Technology Development roject Design Measure against Thermal Failure of RV redicted event progression Relocation of discharged materials The molten core materials discharged through the primary CRGT would be migrated to the lower plenum region. It is necessary to avoid the direct contact of the discharged molten materials with RV (Reactor Vessel). Concept to avoid thermal failure Quenching of molten materials and fragmentation by sodium Effective cooling of fragmented materials ensuring sufficient holding area Inlet plenum Lower plenum Thermal failure of RV RV (Reactor vessel) rimary CRGT Sodium Molten core-material discharge Design measure Quenching of molten material with fragmentation Retention of fragmented material Debris tray RV Concept to avoid the thermal failure of RV 16

融体ジェット崩壊距離 Jet-breakup length, L Lbr br [m] Fast Reactor Cycle Technology Development roject Design Measure against Thermal Failure of RV Design optimization of inlet/lower lenums Quenching of discharged materials No vapor-film formation in FBR condition, being different from LWR condition Rapid fragmentation (jet breakup) due to the thermal interaction with sodium Molten materials LWR condition Coolant Deep penetration of high-temperature jet surrounded by vapor film FBR condition Vapor film Rapid jet breakup due to liquid-liquid contact (No vapor-film formation) 10 8 6 4 Experiments for UO 2 -Na (10m/s) FARO-TERMOS/T1 FARO-TERMOS/T2 Height of lower plenum Maximum jet-breakup length predicted by existing study L br /d j = 2.1(r f /r c ) 0.5 Fr 0.5 where Fr = v j2 /gd j (Saito et al.) 2 Height of 0 lower plenum 0.00 0.05 0.10 0.15 Jet diameter, ジェット直径 d j [m] dj [m] Existing experimental result shows that the achievement of the fragmentation in lower plenum is possible with feasible lower plenum height. Breakup length estimated by past experiments Jet - breakup length 17

Fast Reactor Cycle Technology Development roject Accident scenario in ATWS Initiation of CDA The design measures: Maximum void worth less than ~ 6 $, etc Discharge path Initiating phase No energetics Expansion Structure response Energetics Energetics (re-criticality) Transition phase Gradual progression SA can wall Grid spacer Core Inner duct Design measures: Introduction of FAIDUS Early discharge Large core pool No energetics Material relocation Heat removal Discharge path Cross section Reference for JSFR FAIDUS option 18

Fast Reactor Cycle Technology Development roject Approaches for design and evaluation Basic idea (In Vessel Retention) Utilization of superior coolability of sodium Formation of coolable debris bed by fragmentation and quenching of molten fuel in sodium pool Establishment of coolant circulation by natural convection, two phase cooling inside the debris bed Broaden the fuel debris as much as possible inside the reactor vessel Reduce the amount of the molten fuel which might reach the bottom of RV by upward discharge through inner duct and by in-place cooling inside the core Enlarge the retention capacity by the multilayered debris tray at the bottom of RV 19

Fast Reactor Cycle Technology Development roject Approaches for design and evaluation Core catcher design Multi-layer debris trays Sufficient surface area to retain the core debris as not to exceed coolable and critical limit height Formation of debris bed Fragmentation and dispersion in the lower plenum Debris migration to lower trays by gravity and self-leveling (debris dispersion by the boiling inside the debris) Coolant circulation around the trays Ensure integrity of the trays Guide tubes for debris falling Multi-layer debris tray Debris bed Debris migration to lower tray 20

Fast Reactor Cycle Technology Development roject IVR-Failure Factors in Heat Removal hase Design condition of debris tray It shall hold the amount of core debris corresponding to the best estimate sequence of material relocation phase with some conservative assumptions related uncertainty of debris distribution, composition etc. Capacity is determined as full inventory of core debris (homogeneous debris bed is assumed) Upper lenum 3 to 4min. after shutdown After 1000sec. Upper lenum 20 20 Core Best estimate base fuel distribution Inner 15 10 Outer 40 20 Lower lenum 25 50 Core Lower lenum 21

TEMERATURE( ) Fast Reactor Cycle Technology Development roject 1000 900 800 700 Coolability of the debris bed Coolability of debris bed was confirmed by DEBNET (debris bed temperature analysis coupled with the system code). sodium boiling point DEBRIS DISTRIBUTION DEBRIS DISTRIBUTION on the core support plate on the upper debris plate on the middle debris plate on the lower debris plate 600 500 400 flow reversal of one primary loop 300 0 5000 10000 15000 20000 TIME(sec) Maximum temperature in the debris beds K. Koyama et al, Development of Severe Accident Evaluation Technology (Level 2 SA) for Sodium-cooled Fast Reactors (4) Identification of Dominant Factors in Core Material Relocation and Heat Removal hases, ICA 09, Tokyo, Japan (May 10-14, 2009) No.9126. 22

Fast Reactor Cycle Technology Development roject Conclusions The safety strategy of JSFR against ATWS type CDA implies a controllability of the re-criticality phenomenon reflecting the wide perspective not focusing only on a singular accident phase but looking at the complete set of accident route and scenario. Initiating hase Appropriate core design parameters selected for JSFR are effective in avoiding energetic sequences in the initiating phase. Transition hase The FAIDUS concept is regarded as an effective design measure to eliminate formation of the large molten-fuel pool, which is the key of the transition-phase energetics. 23

Fast Reactor Cycle Technology Development roject Conclusions Material-Relocation hase The event progression prevent re-criticality up to the discharge of molten fuel through CRGT and it is foreseeable to quench the relocated molten-core materials in the lower plenum with sufficient amount of sodium. Heat-Removal hase The JSFR design strategy to enhance the sedimentation-spreading and self-leveling behavior has certain potential to fulfill the requirement of stable core debris cooling, though there remain on-going efforts. 24

Fast Reactor Cycle Technology Development roject Thank you for listening to me today. 25

Fast Reactor Cycle Technology Development roject 26

Fast Reactor Cycle Technology Development roject 1.3 Elements of positive and negative reactivity feedbacks [I. Sato, et al., Elimination of severe recriticality events in the Core Disruptive Accident of JSFR aiming at In-Vessel Retention of the core materials, FR09, Kyoto, Dec., 2009] 27

Fast Reactor Cycle Technology Development roject 2. Review of the established knowledge 2.1 Effective fuel dispersal in voided region Neutron Hodoscope data (CABRI BI3 test) 0 100 200 Time after TO (ms) CABRI Hodoscope Interpretation of the hodoscope data shows that fission gas is released on fuel melting almost instantaneously (τ liq 1ms) and serves as an effective driving pressure for fuel dispersal. [I. Sato, et al., Elimination of severe recriticality events in the Core Disruptive Accident of JSFR aiming at In-Vessel Retention of the core materials, FR09, Kyoto, Dec., 2009] 28

Fast Reactor Cycle Technology Development roject 2.2 High failure threshold of low smear-density fuel Failure Enthalpy (ure TO tests) =F(fission gas retention, smear density) Failure-enthalpy dependency on smear density is also significant under the ULOF condition High failure threshold is expected for JSFR Delayed failure in the un-voided region The LOF condition decreases failure enthalpy ~E liquidus ~E solidus Axially extended molten fuel cavity Failure at high X/L ost-failure fuel relocation in the un-voided region tends to allow negative reactivity feedback. Cladding temperature [I. Sato, et al., Elimination of severe recriticality events in the Core Disruptive Accident of JSFR aiming at In-Vessel Retention of the core materials, FR09, Kyoto, Dec., 2009] 29

Fast Reactor Cycle Technology Development roject 3.4 Simplified expression of the reactivity-competition after fuel motion (simplified expression to clarify dependency on design parameters) 2 Enhanced time delay by high failure threshold 4 Fuel specific heat > 40kW/kg provides early fuel dispersal with enhanced initial fuel temperature and limited disp ρ FCI void 1 Limited ρ void,max Δt disp =f( disp, τ liq ) where disp gives time for fuel melting and τ liq gives time for gas-release from molten fuel 3 Enhanced fuel dispersal reactivity with limited axial core height [I. Sato, et al., Elimination of severe recriticality events in the Core Disruptive Accident of JSFR aiming at In-Vessel Retention of the core materials, FR09, Kyoto, Dec., 2009] 30

Fast Reactor Cycle Technology Development roject 3. Theoretical study on the reactivity balance [H. Niwa, A Comprehensive Approach of Reactor Safety Research Aiming at Elimination of Recriticality in CDA for Commercialization of LMFBR, rogress in Nuclear Energy, Vol.32, No.3/4, pp.621-629, (1998).] 3.1 Void reactivity is always about 2$ when fuel motion starts At fuel motion onset, ρ net approaches 1.0 and increase of ρ void is compensated for by negative ρ Dop and ρ exp ρ net =ρ void +ρ Dop +ρ exp 1 ρ void - 0.7-0.3 1 Fuel motion onset in voided region ρ void Nominal JSFR case ρ net ρ void 2.0 ρ void is always ~2 $ ρ exp ρ Dop ρ fuel motion Time after LOF onset (s) 31

Fast Reactor Cycle Technology Development roject 3.2 Importance of void development at the start of fuel motion r void,max = 5 $ When fuel r void,max = 7 $ motion starts r void» 2$ f void = r void MCVR r void,max Limited ρ void,max gives a larger f void ost-failure competition f void contributes to fuel dispersal (1-f void ) contributes to FCI 32

Fast Reactor Cycle Technology Development roject 4.4 Summary of SAS4A results (2/2) ower history essimistic assumptions tend to raze the power peak. Core-average fuel temperature Slight variation in the temperature history depending on power but all cases are commonly mild without potential of mechanical energy release. [I. Sato, et al., Development of Severe Accident Evaluation Technology (Level 2 SA) for Sodium-cooled Fast Reactors (2) Identification of Dominant Factors in Initiating hase of Unprotected Events, ICA 09, Tokyo, May, 2009] 33

ressure release line Fast Reactor Cycle Technology Development roject 2. Assessments on effectiveness of the design measure 2-2. Application of current knowledge on wall failure Data of out-of-pile and in-pile experiments under EAGLE-1 program are available to consider steel wall failure under sodium cooled condition. Melt transport from EMF Alumina* ~10 kg Inner duct I. D.: 40 mm L: 925 mm Sodium ~ 90 kg Schematic diagram for out-of-pile experiment *) Heating performance of molten alumina for the steel wall is equivalent to that of molten UO 2, since the thermal conductivity of alumina is 2 ~ 3 times higher than that of molten UO 2, which compensate its lower temperature at liquid state. Core center IGR core Test fuel 8 kg-uo 2 Inner duct I. D.: 40 mm L: ~600 mm Sodium ~ 9 kg Schematic diagram for in-pile experiment 34

Fast Reactor Cycle Technology Development roject Two types of steel wall failure were observed under the sodium cooled condition. Steel wall failure without coolant boiling The steel wall failed within a second. Steel wall failure with coolant boiling It took over one hundred seconds until the steel wall failure. 2. Assessments on effectiveness of the design measure 2-2. Application of current knowledge on wall failure Molten fuel Inner duct (2 mmt) Sodium TD8 TD7 VS7 Inner duct failure Molten pool Steel duct Sodium Vapor Heat fluxes from molten material to the wall were evaluated based on temperature increases. mm 150 HV01 100 HT23 50 HT25 0 Inner duct (2 mmt) Molten alumina 20 mm In-pile ID1 experiment data Inner duct failure Out-of-pile IDO-1 experiment data 35

Averaged heat flux [MW/m 2 ] Fast Reactor Cycle Technology Development roject 2. Assessments on effectiveness of the design measure 2-2. Application of current knowledge on wall failure The pool-to-wall heat flux that differentiate the wall failure mode was evaluated. Existence of liquid steel is cause of early wall failure Molten core material is the mixture of fuel and steel. Evaluated heat fluxes to the wall are consistent with SCARABEE BE+3, BF3 and VA tests. The wall failure is dominated by the heat capacity of the wall. Mechanical effect on wall failure should be evaluated in future. Molten pool Early wall failures Sodium (liquid) (within two seconds) Steel wall Wall failure without coolant boiling Wall failure with coolant boiling Vapor Threshold to differentiate wall failure mode Significant delay of the wall failure (over one hundred seconds) Evaluation result The inner duct failure would precede SA can wall failure. 20 15 10 5 WF (gas) FD ID1 WF (Na) ID2 IDO-1 ool-wall (adiabatic) ool-wall (cooled by Na) ool-wall (adiabatic) ool-wall (cooled by Na) Wall-Na UTD-M2c IDO-3 IDO-4 IDO-2 0 0 2 4 6 8 10 Heat transfer duration [s] Fuel-steel mixture Alumina Evaluated average heat flux in the EAGLE-1 program 36

Fast Reactor Cycle Technology Development roject Effect of sodium on fuel discharge Series of out-of-pile and in-pile experiments were conducted under EAGLE-1 program to investigate effects of sodium on fuel discharge. Inside the inner duct At the exit of the inner duct (coolant plenum) Summary of data interpretations are illustrated below. 2. Assessments on effectiveness of the design measure 2-3. Application of current knowledge on fuel discharge Data were obtained in out-of-pile experiments Melt Inner duct Sodium Initial melt ejection through a local opening Sodium vaporization and void expansion rogression of duct failure Massive melt discharge through the voided duct Data were obtained both in out-of-pile and in-pile experiments FCI pressure buildup, but duration of this pressurization was so short that overall discharge did not hinder significantly. 37

TD5 VS4 TD4 VS1 TD3 Fast Reactor Cycle Technology Development roject Temperature [K] Voltage [mv] 80 60 40 20 0 2000 1500 1000 500 VS4 VS1 TD5 TD4 TD3 0 28.5 28.6 28.7 28.8 28.9 29.0 Time [s] 2. Assessments on effectiveness of the design measure 2-3. Application of current knowledge on fuel discharge Inner duct failure Sodium vapor expansion In-pile ID1 experiment data TCs Failure (Fuel arrival) mm Molten alumina 100 HT23 0-925 -980 HT49-1545 HT51 20 mm Inner duct failure ool formation Melt discharge Melt cover gas Sodium in lower plenum Out-of-pile IDO-4 experiment data 38

Fast Reactor Cycle Technology Development roject Blockage formation inside SA (cont.) Bundle effect Since temperature of outer-row fuel-pins are lower than inner-row fuel-pins, fuel dispersion and blockage formation are delayed in sub-channels of outer-row fuel-pins. At first, fuel dispersion and blockage formation occurs in sub-channels of inner-row fuel-pins. Then, penetration of foamy mixture from inner to outer sub-channels Blockage formation inside outer sub-channels Fuel cannot penetrate beyond UAB. Approximately 20 % longer than other subchannel. 2. Assessments on effectiveness of the design measure 2-3. Application of current knowledge on fuel discharge Formation of foamy mixture 1-D fuel motion and blockage formation SA can wall Inner duct Core enetration of foamy mixture into unblocked subchannels Almost all sub-channels are blocked by frozen fuel. Evaluation result The core region can be pressurized both by tight blockage formation and fission gas release from molten fuel. No blockage formation due to bundle effect 39

Fast Reactor Cycle Technology Development roject 2. Assessments on effectiveness of the design measure 2-3. Application of current knowledge on fuel discharge Driving force for upward fuel discharge ressurization of the core region by fission gas due to blockage formation inside SA Release of inter-/intra- granular fission gas due to fuel melting, and formation of foamy (molten fuelgas) mixture. Metallographic observation in CABRI tests Axial dispersion of molten fuel driven by fission gas and blockage formation The blockages were so tight that certain amount of fission gas could not escape through the blockages.» Evidenced by delayed gas escape from the disrupted region measured by void detector and bell jar CABRI T2 test 40

Fast Reactor Cycle Technology Development roject Upward fuel discharge through the inner duct Fission gas released from melting fuels forms foamy (molten fuel-gas) mixture. Metallographic observation in CABRI experiments Continuous supply of fission gas by axial progression of fuel melting before onset of fuel discharge Discharge of fission gas toward the opening section (the upper end of the inner duct) should accompany molten fuel Optical observation of STAR experiment henomenological consideration based on a scientific book of the two-phase flow Confirmation of fuel discharge through the sodium-filled path Negligible effect of sodium on fuel discharge confirmed by out-of-pile and in-pile experiments in EAGLE program 2. Assessments on effectiveness of the design measure 2-3. Application of current knowledge on fuel discharge ressurization by fission gas release and blockage formation No easy separation of fission gas from molten fuel No blockage formation and rapid discharge Limited slip ratio between molten fuel and gas Core SA can wall Close end Inner duct 41

Fast Reactor Cycle Technology Development roject Upward fuel discharge through the inner duct (cont.) Sodium vapor supplied from the lower part of the inner duct can become the driving force. Observed in MELT experiments. Data of upward discharge experiment in EAGLE-2 program also suggest effect of sodium vapor on upward discharge. Fission gas blow from the fuel pin gas plenum after depressurization of core region can also become the driving force. 2. Assessments on effectiveness of the design measure 2-3. Application of current knowledge on fuel discharge Sodium vapor supply Evaluation result Sufficient driving force for upward fuel discharge through the inner duct would be obtained, and fuel would be discharged rapidly. Close end Fuel pin gas plenum 42

Ratio of fuels remaining in each region [-] Normalized power [/ 0 ] Ratio of fuels remaining in the core [-] Net reactivity [$] Fast Reactor Cycle Technology Development roject The inner duct failure preceded SA can wall failure, and molten fuel discharged through the inner duct. Can wall failed after cease of fuel discharge. The higher the thermal output was, the more fuel discharged. Distribution of fuel remaining in the core was given as the initial condition of the subsequent phase. Negative reactivity insertion by fuel discharge reduced the thermal output of the core effectively. Cease of progressive fuel melting 1 0.8 0.6 0.4 0.2 0 1 0.95 0.9 0.85 0.8 0.75 0.7 1000 100 10 1 Region 1 Region 2 Region 3 Region 4 Region 5 Region 6 Region 7 Total 0.1 18 19 20 21 22 23 24 25 Time from LOF onset [s] 3. Fuel Discharge Analyses using the SIMMER-III code 3-2. Results 1 2 3 40 5 6 7 Region No. No fuel discharge CW failure 2 0-2 -4-6 -8-10 -12-14 43

Ratio of fuels discharged from the core [-] Fast Reactor Cycle Technology Development roject 3. Fuel Discharge Analyses using SIMMER-III code 3-2. Results Core-wide fraction of molten fuel is less with earlier fuel discharge. Criterion for amount of discharged fuel Determined by the evaluated reactivity change 0.4 during the subsequent phase in terms of 0.35 avoiding recriticality No requirement in terms of preventing of 0.3 molten pool formation Uncertainties in evaluation of inner duct failure Heat-transfer from the molten core materials to the steel wall Effect of mechanical load on wall failure. It is planned to consolidate the evaluation methodology for the inner-duct failure in the future R&D. 0.25 0.2 0.15 0.1 0.05 Inner duct failure and initiation of fuel discharge terminates progressive fuel melting. The earlier fuel discharge initiates, the less discharged fuel amount is. Uncertainty predicted by EAGLE-1 program 0 44 45 46 47 48 49 50 Integrated power of core until termination of fuel discharge [FS] 44

Fast Reactor Cycle Technology Development roject Appendix Experimental database (1/2) Table. Experimental databases for discharge path formation The element of phenomenological diagram Experimental database Release of inter-/intra- granular fission gas Sodium vapor Heat transfer from molten core materials Wall coolability Blockage inside SA Bundle effect CABRI-1 tests CABRI-2 tests TREAT: L5, L6, L7, L8, LO-series EAGLE-1 (Out-of-pile / In-pile) SCARABEE: BE+3, V-A, BF3 CABRI-2: E2, E3, E13, E11 CABRI-RAFT: T2, T-A1 LANDTL (JAEA) Table. Experimental databases for discharge (1/2) The element of phenomenological diagram Fuel pin plenum gas release Blockage inside SA CABRI tests Experimental databases CABRI-2: E2, E3, E13, E11, CABRI-RAFT: T2, T-A1 SIMBATH MELT(JAEA), Geyser (CEA), Blocker (JRC Ispra), THEFIS(KIT) LANDTL (JAEA) 45

Fast Reactor Cycle Technology Development roject The element of phenomenological diagram Appendix Experimental database (2/2) Table. Experimental databases for discharge (2/2) Experimental databases Release of inter-/intra- granular fission gas Sodium vapor Sodium vapor (After the inner duct failure) Gas separation from molten core material Liquid-gas slip ratio in the discharge path FCI Heat transfer to sodium Heat transfer to the wall CABRI-1 tests, CABRI-2 tests TREAT: L5, L6, L7, L8, LO-series MELT (JAEA) EAGLE-2 (Out-of-pile / In-pile) CAMEL-C6, C7 (ANL) CABRI-2: E2, E3, E13, E11, CABRI-RAFT: T2, T-A1 b-bi pool experiment (Kyoto Uni., JAEA) EAGLE-2 (Out-of-pile / In-pile) MELT (JAEA) STAR experiments EAGLE-1 (Out-of-pile) MELT (JAEA) THINA (KIT), FARO-TERMOS (JRC Ispra) FRAG (SNL), M-series (ANL) EAGLE-1 program (Out-of-pile / In-pile) MELT (JAEA) EAGLE-1 program (Out-of-pile / In-pile) MELT(JAEA), Geyser (CEA), Blocker (JRC Ispra), THEFIS(KIT) 46

Fast Reactor Cycle Technology Development roject 3. Design Measure against Severe Energetics 3-1. redicted event progression (2/5) Falling of absorber rod (B 4 C) Installation of SASS utilizing temperaturesensing alloy in backup RSS Elimination of unsuccessful causes (unreach of heated coolant etc.) during CDA No holding force of electromagnet at 800 o C Iron core Coil Electromagnet part Control rod driving mechanism RV head Driving shaft Expected convection enhanced by power peaking in the initiating phase SASS Iron core Temperaturesensing alloy (Ni-Co-Fe) Connecting surface Armature part (Absorber rod) SASS Coolant inlet part Electromagnet part Connecting surface Armature part Control rod Core fuel Holding state Detached state Nakanishi, et al., Nucl. Tech., 170, No. 1 (2010). Evaluation result Absorber rod in backup CRGT should be falling at the beginning of material-relocation phase. 47

Fast Reactor Cycle Technology Development roject 2. Approaches for design and evaluation In-Vessel Retention is more rational than Ex-Vessel Retention In Vessel Retention Merits: Experimental data and evaluation models are available for debris bed cooling, some data is available for fragmentation and quenching Less additional plant material (only internal core catcher) Demerits: Additional structure (internal core catcher) inside RV might cause unexpected trouble, load of inspection. Ex Vessel Retention Merits: Less impact on RV design Demerits: Larger uncertainty in relocation process and cooling capability, additional phenomena to be considered, i.e., sodium-concrete reaction, hydrogen burning. Larger additional plant material, i.e., external core catcher, cooling circuit, steel liner and thermal insulator etc. 48

Fast Reactor Cycle Technology Development roject 6. Integrity of debris trays 6-1. Review of past design Review of past design SUER HENIX 1 SNR-300 Large LMFBRS evaluated by GE UK LMFBRS (DFR, FR, CDFR) DFBR SUER HENIX 1 Key phenomena related to the integrity of debris trays Fragmentation of molten core debris within the lower plenum Cooling by natural circulation of the sodium in the reactor vessel Considering these phenomena, debris trays design works has been conducted to remain the integrity against whole core debris accumulation. 49

Fast Reactor Cycle Technology Development roject 6. Integrity of debris trays 6-2. Status of design works Status of design works Design optimization of the lower plenum To ensure the depth and the amount of sodium for the quenching of molten debris Consideration of debris trays design To enhance debris dispersion To keep the debris-bed thickness within each section To restrain the thermal deformation To reduce the influence of coolant flow from C/L pipes during normal operation Future lan Evaluate the integrity under CDA conditions and normal operation Determine detailed debris trays design with safety requirements Depth of the lower plenum Multi-layer debris trays 50

Fast Reactor Cycle Technology Development roject rotection against direct melt jet attack 6. Integrity of debris trays 6-2. Status of design works Aimed at In Vessel Retention utilizing higher cooling capability of liquid sodium Multi-layered Reactor core Debris Tray for debris retention within limit bed height of cooling and sub-critical state core support structure Enlargement of Coolant Inventory for molten fuel quenching and fragmentation into small particles Chimney Reactor vessel lower structure for effective coolant circulation Guide Tubes for fuel debris settling to lower plate debris goes down when the bed height exceeds certain level by fluidization 51

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-1. Sedimentation Available knowledge Debris size and shape, configuration of debris bed: Melt injection into liquid pool (FARO, KROTOS, EAGLE etc) Heap bed (0.22m high, 0.5m radius) is assumed in SX based on a sedimentation experiment. Some experimental and theoretical studies are available in the field of geographic science about transportation process of particles and sedimentation. Angle of repose : physical property of particle depends on particle size, density, shape, ambient etc. 52

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-1. Sedimentation Falling process of particle cloud (1/2) A correlation indicates that deposit diameter is 1/2 of fall height (D=Z/2). In our case, Zf is in order of 10m. Ring deposit mode is expected. D>Z/2 due to mild FCI at melt injection. Observed in grass spheres injection into water pool Thermal behaves as particle cloud. Swarm as independent particles. Zf: Fall out height John W. M. Bush et al., article clouds in homogeneous and stratified environments, J. Fluid Mech. (2003), vol. 489, pp.29-54 53

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-1. Sedimentation Falling process of particle cloud (2/2) Some prediction models have been developed in the field of coastal civil engineering for land reclamation using hopper barges. It was found that for deeper pool sediment becomes flatter. Flatter sediment is expected for JSFR case. ool depth shallow deep mountain trapezoid ring Sediment shape dispersed Analysis for stone rubble/water Oda et al (1993): Coastal Engineering (in Japanese), vol. 40, pp.951-955 stone/water MOX/Na article density (ρp) kg/m 3 2650 9980 Liquid density (ρ) kg/m 3 1000 735 Liquid viscosity (ν) m 2 /s 1.00E-06 3.32E-07 article radius (a) m 1.5E-03 2.5E-04 Numer of particle (Np) - 3.5E+03 9.6E+07 article terminal velocity (ws) m/s 3.8E-01 4.3E-01 article Reynolds number (Rep=ws*a/ν) - > 500 > 500 Froude nember (Fr=ws^2/Hg) ratio - 1 < 1 54

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-1. Sedimentation Status (reliminary Stage) Homogeneous distribution of discharged fuel through holes with certain pitch (like a shower nozzle) B Dispersion in the falling process in the lower plenum, FCI, flow turbulence and B particle cloud behavior. Transportation to the lower layers through debris guide tubes. B B Rad. Blanket Fuel discharge path in bottom plate of CSS In case of corn shape sedimentation, peak height is limited by angle of repose. Self leveling B B B B (CSS) B Inlet plenum Lower plenum B B B Inner Core B B B B B Outer Core Core Barrel Rad. Shield 55

Bed Maximum Height (m) Bed Radius (m) 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0 Fast Reactor Cycle Technology Development roject 30% fuel and 30% steel, bed porosity 50% Dry out height 0 0 10 20 30 40 50 Angle (Degree) Number of discharge holes 10 20 30 Angle of Repose? 7. Debris bed formation 7-1. Sedimentation 2.5 2 1.5 1 0.5 Status (reliminary Stage) If ideal corn shape sedimentation is assumed, the peak height is over the dry out height. Quantitative evaluation will be required for sedimentation process. 1/2H Angle H Sandglass effect Maximum height is reduced to ½ Transport time is several ten seconds (10cm hole) 56

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-2. Self leveling Available knowledge (early 1980s) Self leveling: flattening of debris bed surface due to local coolant boiling ANL bed leveling experiments; Induction heating, UO2-steel mixture in water SNL In-pile experiments (D-series) in Sodium Status(Up-to-date) Kyushu University, JAEA; rediction model for self leveling onset is proposed, which shows good agreement with an experiment (depressurization, bottom heating, alumina, zirconia, stainless steel, lead, in water) It is estimated that self leveling can be set off under reactor condition. Some uncertainty in the rate of self leveling. 57

Fast Reactor Cycle Technology Development roject Example of the experimental results 7. Debris bed formation 7-2. Self leveling 0 s 150 s 300 s 450 s 1 mm, Spherical, Alumina Bin ZHANG et al., Self-Leveling Onset Criteria in Debris Beds, J. Nucl. Science and Tech., Vol.47, No.4, p.384-395 (2010) 58

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-2. Self leveling rediction for reactor case The prediction model for self leveling is based on a force balance equation for a single spherical particle. It is shown that the self leveling is possible to happen in the debris bed more than 10cm in height. Onset criteria V g -V eq >0 F D +F b =F g Assumed condition V eq is a function of F D V g is a function of bed power density and height V eq :relative velocity between particle and two phase flow V g : two phase flow velocity F D :drag F b :buoyancy F g :gravity Bin ZHANG et al., Self-Leveling Onset Criteria in Debris Beds, J. Nucl. Science and Tech., Vol.47, No.4, p.384-395 (2010) 59

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-3. Future plan Sedimentation Experiments; particle cloud injection into water pool, molten material injection into sodium pool Model development for evaluating sedimentation radius and accumulated configuration and its application Determination of design condition of debris tray Self leveling Extension of experimental data and model development Code development for transient process evaluation of self leveling and its application 60

Fast Reactor Cycle Technology Development roject 8. Conclusions Approaches In-Vessel multi-layered debris tray is introduced in RV design Utilization of superior thermal property of sodium, formation of debris bed Main factors for IVR failure in the heat removal phase No thermal failure of RV caused by the debris tray failure Status of evaluation Limit height of debris bed (coolability, criticality) is identified. Coolability of the debris bed can be maintained subsequent conditions to the material relocation scenario. Integrity of debris trays is under evaluation. Formation process of debris bed can be evaluated from the view points of sedimentation and self leveling. Future plan is identified. 61

Fast Reactor Cycle Technology Development roject 5. Coolability of the debris bed 5-1. Evaluation conditions Material distribution based on a material relocation scenario Case1: Upward ejection in Transition hase to cause shutdown Case2: Downward ejection of fuel through CRGT Case3: Whole fuel accumulates on the core catcher (bounding) Debris composition and heat generation (timing) Case1: 20% of fuel and cladding steel (0 s) Case2: ~50% of fuel and cladding steel (100 s) Case3: 100% of fuel, 50% of axial blanket and cladding steel (1000 s) 70% of decay heat is accounted (volatile Fs escape ) Debris bed parameters (BE base) homogeneous, 0.3mm diameter, 50% porosity Cooling condition: Natural circulation of DHRS Success criteria: No Dryout in debris beds (& Structural integrity) 62

Temperature ( ) Fast Reactor Cycle Technology Development roject 5. Coolability of the debris bed 5-2. Evaluation results (1) 1000 900 800 700 600 500 Case1: Upward ejection in T Max. Temp. in Debris Bed No coolant boiling Envelopes no dryout criterion Max temp. of structure below the debris bed ~800 (Adiabatic assumption at bottom ) Up to 40% of fuel inventory could be cooled without dryout 400 300 0 5000 10000 15000 20000 Time (s) 63

Temperature ( ) Fast Reactor Cycle Technology Development roject 5. Coolability of the debris bed 5-3. Evaluation results (2) Case2: ~50% of fuel through CRGT 1000 900 800 700 600 Max. Temp. in Debris Bed No coolant boiling Max mid-wall temp. of structure below the debris bed ~708 (Downward heat removal considered) 500 400 300 0 5000 10000 15000 20000 Time (s) 64

Temperature ( ) Fast Reactor Cycle Technology Development roject 5. Coolability of the debris bed 5-4. Evaluation results (3) Case3: 100% of fuel and 50% of blanket 1000 900 800 700 Max. Temp. in Debris Bed No coolant boiling Max mid-wall temp. of structure below the debris bed ~710 (Downward heat removal considered) 600 500 400 300 Flow reversal of a loop 0 5000 10000 15000 20000 Time (s) 65

Fast Reactor Cycle Technology Development roject B 5. Coolability of the debris bed 5-5. Evaluation results (4) RV SG B Flow reversal of a loop DHRS IHX Normal operation Smaller heat removal Larger heat removal Damaged core B B Operation of blower of a loop to force imbalance as an AM Reversed flow direction Blocked Debris bed Because of unstable configuration (Core almost blocked, heat source at the bottom of RV), a slight imbalance in two DHRS loops cause the flow reversal in a loop of two. Consequent loop flow increases contributes to long term cooling. 66

COOLANT FLOW RATE(kg/s) COOLANT FLOW RATE(kg/s) Fast Reactor Cycle Technology Development roject 5. Coolability of the debris bed 5-6. Evaluation results (5) 400 200 0-200 1-RACS HEAT REMOVAL RATE 100% 99% 95% 90% 85% 80% 50% 0% Larger the imbalance, earlier the flow reversal -400-600 -800-1000 0 1000 2000 3000 4000 5000 6000 TIME(sec) arametrically reducing heat removal of a RACS for imbalance between DHRS loops 1200 1000 800 600 400 200 0-200 -400-600 -800-1000 ordinary flow reverse flow 0 5000 10000 15000 20000 TIME(sec) 67

Fast Reactor Cycle Technology Development roject 5. Coolability of the debris bed 5-7. Evaluation summary Coolability of debris bed was confirmed by debris bed temperature analysis coupled with the cooling system, according to the following material relocation scenario Case1: Upward ejection in Transition hase to cause shutdown Case2: Early downward ejection of fuel through CRGT Case3: Whole fuel accumulates on the core catcher (bounding) The flow reversal of a primary coolant loop of the two loop system of the JSFR which is caused by possible imbalance between two DHRS loops increase the flow in RV. Helpful for long-term cooling. 68

Fast Reactor Cycle Technology Development roject 7. Debris bed formation 7-2. Self leveling 69

Fast Reactor Cycle Technology Development roject 0 sec 10 sec 30 sec 50 sec 0.5 mm,sphere,alumina,q=0.21 W/cc 0 sec 20sec 50 sec 80 sec 0.5 mm,sphere,zirconia,q=0.21 W/cc 70

Fast Reactor Cycle Technology Development roject 0 sec 50 sec 100 sec 150 sec 0.41 mm,non-sphere,ss,q=0.22 W/cc 71