Super LWR (SCWR) Conceptual Design Yoshiaki Oka Professor The University of Tokyo UT-UCB Internet seminar December 8, 2006

Generation IV Reactor Concepts SCWR (Super LWR / Super FR) SFR LFR VHTRGFR MSR 3

Features of Super LWR/Super FR Simple & compact plant systems –No water/steam separation –Low flow rate, high enthalpy coolant High temperature & thermal efficiency (500C, ～ 44% ） Utilizations of current LWR and Supercritical FPP technologies –Major components are used within the temperature range of past experiences Supercritical FPP （ once-through boiler ） Super LWR/ Super FR BWRPWR

Evolution of boilers and supercritical fossil fired power plant technologies

Circular Boiler Water tube boiler Once-through boiler LWR Super LWR (SCWR ) Evolution of boilers

Supercritical Water in the Power Industry Most new coal-fired power plants are supercritical. Coal-fired SC plants in the world and their performance Source: World Bank Organization

Toshiba: 700 MWe (24MPa, 593/593°C) SC turbines are proven technology MHI: 1000 MWe (24.5MPa, 600/600°C) Major vendors of SCW components include GE, Toshiba, Hitachi, MHI, B&W, Siemens

Density and specific heat of supercritical water (25 MPa)

Control rods Supercritical water TurbineGenerator Condenser Pump Heat sink Reactor Core 280 ℃ 500 ℃ Super LWR Super LWR: Supercritical-Water-Cooled and Moderated Reactor developed at Univ. of Tokyo Once-through direct cycle thermal reactor Pressure: 25 MPa Inlet: 280 ℃ Outlet (average): 500 ℃ Flow rate: 1/8 of BWR

Scope of studies and Computer codes 1.Fuel and core Single channel thermal hydraulics (SPROD), 3D coupled core neutronic/thermal-hydraulic (SRAC- SPROD), Coupled sub-channel analysis, Statistical thermal design method, Fuel rod behavior (FEMAXI- 6), Data base of heat transfer coefficients of supercritical water 2. Plant system; Plant heat balance and thermal efficiency 3. Plant control 4. Safety; Transient and accident analysis at supercritical-and subcritical pressure, ATWS analysis, LOCA analysis (SCRELA) 5. Start-up (sliding-pressure and constant-pressure) 6. Stability (TH and core stabilities at supercritical and subcritical-pressure) 7. Probabilistic safety assessment

Fuel and Core design

Core design criteria Thermal design criteria Maximum linear heat generation rate (MLHGR) at rated power ≦ 39kW/m Maximum cladding surface temperature at rated power ≦ 650C for Stainless Steel cladding Moderator temperature in water rods ≦ 384C (pseudo critical temperature at 25MPa) Neutronic design criteria Positive water density reactivity coefficient (negative void reactivity coefficient) Core shutdown margin ≧ 1.0%ΔK/K

Fuel assembly design Design requirementsSolution Low flow rate per unit power (< 1/8 of LWR) due to large ⊿ T of once-through system Narrow gap between fuel rods to keep high mass flux Thermal spectrum coreMany/Large water rods Moderator temperature below pseudo-critical Insulation of water rod wall Reduction of thermal stress in water rod wall Uniform moderationUniform fuel rod arrangement UO 2 + Gd 2 O 3 fuel rod UO 2 fuel rod Control rod guide tube Water rod ZrO 2 Stainless Steel Kamei, et al., ICAPP’05, Paper 5527

Fuel enrichment 4.2m 2.94m 1.26m 5.9wt%UO 2 6.2wt%UO 2 4.2m 2.94m 1.26m 5.9wt%UO wt%Gd 2 O 3 6.2wt%UO wt%Gd 2 O 3 (a) UO 2 fuel rod(b) UO 2 + Gd 2 O 3 fuel rod Fuel enrichment is divided into two regions to prevent top axial power peak Average fuel enrichment 6.11wt%

Coolant flow scheme Mix Inlet: Outlet: Outer FA Inner FA CR guide tube CoolantModerator Inner FAUpwardDownward Outer FADownward Flow directions To keep high average coolant outlet temperature Kamei, et al., ICAPP’05, Paper 5527

3D Coupled Core calculation Neutronic calculation: SRAC code system (JAERI) Thermal-hydraulic calculation: SPROD (Univ. Tokyo) 3D Core burnup calculation （ 1/4 core ） FA FA segment q c (i) q w (i) pellet cladding coolant moderator Water rod wall Single channel thermal- hydraulic analysis （ SPROD) Average condition Calculation mesh 2-D FA burnup calculation （ 1/4 FA ） SRAC and ASMBURN Macro cross section N -TH coupling

3-D N-T Coupled Core Calculation T-H calculation based on single channel model Neutronic calculation; SRAC Core consists of homogenized fuel elements Fuel assembly Homogenized Fuel element 1/4 symmetric core Single channel T-H analyses 3-D core calculation q c (i) q w (i) pellet Cladding Coolant Moderator Water rod wall Single channel T-H model

Fuel load and reload pattern 1 st cycle fuel 2 nd cycle fuel 3 rd cycle fuel 4 th cycle fuel (a) 1 st → 2 nd cycle(b) 2 nd → 3 rd cycle(c) 3 rd → 4 th cycle ¼ symmetric core 120 FAs of 1 st,2 nd and 3 rd cycle fuels and one 4 th cycle FA 3rd cycle FAs which have lowest reactivity are loaded at the peripheral region of the core to reduce the neutron leakage This low leakage core is possible by downward flow cooling in peripheral FAs

Coolant flow rate distribution Relative coolant flow distribution (1/4 core) Flow rate to each FA is adjusted by an inlet orifice to match radial power distribution 48 out of 121FAs are cooled with descending flow FA with ascending flow cooling FA with descending flow cooling

Control rod patterns X : withdrawn rate (X/40) Blank box : complete withdrawal (X=40) At the EOC, some CRs are slightly inserted to prevent a high axial power peak near the top of the core Prevent a high MCST

Power distribution 2.2 ～ 3.3GWd/t7.7 ～ 8.8GWd/t13.2 ～ 14.3GWd/t Radial power distribution (¼ core) Radial power peaking factor is (Only Upward flow cooling region ) Axial power peaking factor is Axial power distribution

Coolant core outlet temperature and Maximum cladding surface temperature distribution (a) Coolant outlet temperature distribution (1/4 core) (b) Maximum cladding surface temperature distribution (1/4 core) Coolant temperature of inner FA is C (average 500C) Coolant temperature of peripheral FA is C

MLHGR and MCST MLHGR and MCST are kept below 39kW/m and 650C throughout a cycle respectively Thermal design criteria are satisfied (a) MLHGR(b) MCST

Water density reactivity coefficient and Shutdown margin Water density reactivity coefficient is positive (negative void reactivity coefficient) Shutdown margin is 1.27 %dk/k All CR clusters are inserted except the maximum worth cluster Fuel and coolant temperature are 30C No Xe or other FP in the core Neutronic design criteria are satisfied

Super LWR characteristics summary CoreSuper LWR Core pressure [MPa]25 Core thermal/electrical power [MW] Coolant inlet/outlet temperature [C]280/500 Thermal efficiency [%]43.8 Core flow rate [kg/s]1418 Number of all FA/FA with descending flow cooling121/48 Fuel enrichment bottom/top/average [wt%]6.2/5.9/6.11 Active height/equivalent diameter [m]4.2/3.73 FA average discharged burnup [GWd/t]45 MLHGR/ALHGR [kW/m]38.9/18.0 Average power density [kW/l]59.9 Fuel rod diameter/Cladding thickness (material) [mm]10.2/0.63 (Stainless Steel) Thermal insulation thickness (material) [mm]2.0 (ZrO 2 )

Sub-channel analysis coupled with 3D core calculation

Principle for Preventing Cladding Failures Super LWR: no boiling, limit cladding temperature BWR, PWRSuper LWR Normal operation Sufficient margin to BT No creep rupture 1) (Design limit temperature for normal operation) Abnormal transient No BTNo plastic strain & no buckling collapse 2) (Design limit temperature for abnormal transient) 1) A. Yamaji, Y. Oka, J. Yang, et al., “Design and Integrity Analyses of the Super LWR Fuel Rod.,” Proc. Global2005, Tsukuba, Japan (2005) 2) A. Yamaji, Y. Oka, Y. Ishiwatari, et al., “Rationalization of the Fuel Integrity and Transient Criteria for Super LWR,” Proc. ICAPP’05, Seoul, Korea (2005) Accurate evaluation of the peak cladding temperature is essential

Characteristics of supercritical water cooling Large axial Δρ ⇒ induces flow distribution within FA High sensitivity of (ΔT/ΔH) near the core outlet Outlet Inlet Subchannel analysis is required to accurately determine Maximum Cladding Surface Temperature (MCST) 3-D core calculation: –T-H calculation based on single channel model –Underestimated MCST

Overpower/ Over temperature (abnormal transients) Applicable radial and axial flux factor Applicable local flux factor Engineering uncertainties Nominal steady state core average condition 25MPa, outlet 500 ℃... etc Nominal peak steady state condition Maximum peak steady state condition (95/95) 3-D core calculations Subchannel analyses Statistical thermal design Limit for design transients (95/95) Plant safety analyses Margin Failure limit Evaluation of Max. peak cladding temperature Yamaji, et al., Global 2005, Paper 556 DNBR 1.0 (95/95) → 1.17 DNBR 1.3 DNBR ℃

Method: Calculation Flow N&T-H coupled subchannel analyses Pin power distribution f(burnup history, density, CR insertion) 3-D core calculations (Homogenized FA model) Core power, burnup, density, distribution CR patterns Reconstructed pin power distribution Subchannel analysis Peak cladding surface temperature

Subchannel Analysis Code Subchannel analysis code for Super LWR FA developed at the Univ. of Tokyo 1) 3 type of subchannels –Type A: corner of WR –Type B: side of WR –Type C: corner of FA WR thermally insulated (ZrO 2 ) Heat transfer correlation: Watts (1) T. Tanabe, S. Koshizuka, Y. Oka, et al., “A subchannel analysis code for Supercritical-pressure LWR with Downward-flowing water rods,” Proc. ICAPP’04, Pittsburgh, PA, USA (2004) 2.5% neutron heating assumed

Basic Characteristics of Super LWR FA Fuel rod number Relative fuel rod power Coolant temperature Power Temperature Type “A” Type “B” Type “C” Type Pin power (Thermal neutron flux) Heat transfer to WR (Wetted length/ Heated length) ALowSmall (1.43) BHighLarge (1.70) CMidV.Large (2.53) 3 types of subchannels and fuel rods Uniform outlet temperature

Reconstruction of pin power distributions Height [m] Normalized power Core power distributions (3-D core calculations) Pin power distribution f(burnup history, density, CR insertion) Homogenized FA Reconstructed pin power distribution Coupled subchannel analyses

[kg/m 2 sec ] [℃][℃] Mass Flux and MCST Distributions [kg/m 2 sec] FA-a2 (Large gradients) [kg/m 2 sec ] [℃][℃] FA-b (Gadolinia rods) FA-c (CR withdrawal) [℃][℃] Mass flux distributions MCST distributions

Comparison of Single Channel Analysis and Subchannel Analysis MCST predicted by subchannel analysis is about 58 ℃ higher than that predicted by single channel analysis Large power gradient inside FA should be avoided by design FA-a1 (large power gradient) FA-a2 (large power gradient) FA-b (Gadolinia rods) FA-c (CR withdrawal at EOC) 44 ℃ 58 ℃ 49 ℃ 46 ℃ MCST ΔT from single channel analysis to subchannel analysis

Peak Cladding Surface Temperature Nominal steady state core average condition Nominal peak steady state condition Maximum peak steady state condition 3-D core calculations Subchannel analyses Statistical thermal design Limit for design transients Plant safety analyses Failure limit Nominal peak steady state condition (Homogenized FA) (ΔT 1 =150 ℃ ) (ΔT 2 =58 ℃ ) (ΔT 3 =?) (ΔT 4 = ? ) Ave. outlet:500 ℃ 708 ℃ Peak cladding surface temperature Criterion: ? ? ? 650 ℃

Statistical Thermal Design Taking uncertainties into evaluation of peak cladding temperature

Methods to evaluate the engineering uncertainty  Classification: (1) The direct method: All uncertainties are set at their worst values and occur at the same location and at the same time. Traditional and conservative. (2) The traditional way by using hot spot and hot channel factors: (a) The deterministic method by using factors. (b) The statistical method by using factors. (c) The semi-statistical method: Two groups of uncertainties: direct and statistical factors. The factors are evaluated separately and combined statistically. (3) The statistical thermal design method: System parameters uncertainties are combined statistically. Uncertainties of nuclear hot factors are considered statistically. Engineering hot spot factors are used in a statistical way.

Uncertainties considered in statistical design procedure system parameter uncertainties Inlet coolant temperature uncertainty Core power uncertainty Inlet coolant flow rate uncertainty Core pressure uncertainty Uncertainty of the ratio of the flow rate in water rods to the whole flow rate hot factor uncertainties Nuclear hot factor Engineering hot spot factor Heat transfer correlation uncertainty

Nuclear and engineering hot factors Engineering hot spot factors are used to consider some engineering uncertainties, such as manufacturing tolerances, data uncertainties, etc. Nuclear hot factors are used to consider the power distribution in the hot assembly, in the hot channel and at the hot spot. The nuclear hot spot factor. The ratio of the maximum linear heat flux to the average linear heat flux in the core The axial nuclear hot assembly factor. The Ratio of the max planar power to the average planar power in the hot assembly The radial nuclear hot assembly factor. The ratio of the hot assembly power to the average assembly power DefinitionFactors The cladding surface temperature rise engineering hot spot factor The coolant temperature rise engineering hot spot factor DefinitionFactors

Engineering sub-factors Sub-factorsUncertainties explained by sub-factors Nuclear dataNuclear properties of the fuel rods Power distributionShapes of power distribution in hot assembly Fissile fuel content tolerancesEnrichment and amount of fissile material Inlet Flow maldistribution Assembly hydraulic resistance and orifice uncertainties Flow distribution calculationIntra-assembly flow maldistribution Subchannel flow area Geometric tolerances of fuel rod diameter and pitch on sub-channel flow area Pellet-cladding eccentricityEccentric position of fuel pellet within cladding Coolant propertiesData of coolant properties Cladding propertiesThickness and thermal conductivity of cladding Gap propertiesConductivity of gap between fuel and cladding

 Monte Carlo statistical procedure Step 1: To sample all parameters and hot factors uncertainties following their distributions and to combine them into groups. Step 2: For each group of samples, sub-channel code is used to calculate the MCST of the core. Step 3: The MCST distribution is analyzed to get the standard deviation of system parameters and hot factors. The correlation uncertainty is evaluated. Total uncertainty is: Step 4: The engineering uncertainty is evaluated as:

 Monte Carlo statistical procedure Engineering uncertainty is evaluated as: k=1.645 is to ensure 95/95 limit.

Sub-channel area uncertainty - Fabrication tolerance, Cladding strain, etc. Monte Carlo Process 5000 groups of samples of locations and diameters of all fuel rods in the 1/8 hot assembly Calculation conditions

Results of MCST distributions by Monte Carlo procedure Distributions of MCST of case 1 (Left) and case 2 (right) Case 1: system parameters are sampled as normal distributions Case 2: system parameters are sampled as uniform distributions

Statistical characteristics of MCST distributions Case 1: system parameters are sampled as normal distributions Case 2: system parameters are sampled as uniform distributions

Engineering uncertainty: Standard deviation of system parameter uncertainty and hot factor uncertainty Standard deviation of correlation uncertainty Thermal margin for engineering uncertainty

Peak Cladding Surface Temperature Nominal steady state core average condition Nominal peak steady state condition Maximum peak steady state condition 3-D core calculations Subchannel analyses Statistical thermal design Limit for design transients Plant safety analyses Failure limit Nominal peak steady state condition (Homogenized FA) (ΔT 1 =150 ℃ ) (ΔT 2 =58 ℃ ) (ΔT 3 =32 ℃ ) (ΔT 4 = ? ℃ ) Ave. outlet:500 ℃ 708 ℃ Peak cladding surface temperature Criterion: ? ℃ 740 ℃ ? ℃ 650 ℃

Mechanical strength requirements for fuel cladding at normal operating condition

FEMAXI-6 Fuel Analysis Code Models Fuel analysis code for LWR Thermal and mechanical coupled analysis Temperature, FP gas release, FEM calculations for each segment SUS304 for cladding Irradiation creep and corrosion by oxidation are not calculated

Time to rupture [h] Creep rupture strength [MPa] 750 ℃ 700 ℃ 650 ℃ 600 ℃ ℃ 650 ℃ 700 ℃ 750 ℃ PNC1520PNC316 Basic Fuel Rod Behavior Peak cladding centerline temperature: 757 ℃ at Segment No. 9 (peak cladding surface temperature: 740 ℃ ) Creep rupture strength becomes most limiting at high temperature Primary membrane stress needs to be evaluated Compressive (BOL) ⇒ tensile (EOL) due to PCMI Primary membrane stress on cladding Creep rupture strength of advanced SS (JNC) Yamaji, et al., Global 2005, Paper 556

Fuel rod designs and cladding hoop stresses Placing gas plenum in lower part of fuel rod is effective in reducing stress (required gas plenum volume reduced to 1/5) Constraints on P internal Gas plenum position Gas plenum / fuel vol. ratio Clad hoop stress (at seg. No. 9) [MPa] Less than P coolant Upper part ～ ～ +100 Lower part ～ ～ +59 No constraints Upper part ～ ～ +68 Lower part ～ ～ +36 (“Zero PCMI” design) SCWR IEM Tsukuba, Oct.12, 2005

Mechanical Strength Requirements for Fuel Rod Cladding (at normal operation) Requirements: –Life time ： 36,000 ～ 48,000 hours （ 500days×3 ～ 4 ） –Peak temperature ： 757 ℃（ centerline ） / 740 ℃ (surface) Requirements exceed mechanical strengths of commercially available stainless steels (e.g. SUS316) Advanced Austenitic Stainless Steel of JNC （ PNC1520 ） almost meets the requirement - Creep rupture strength of approx. 90MPa (750 ℃, 10,000hours) (plenum/fuel) vol. ratio Plenum positionConstraint on P internal Stress 0.1BottomNo constraints-64 to BottomLess than P coolant -96 to +100 Primary membrane stress on the cladding

Safety

Super LWR BWRPWR Once-through cooling ⇒ No circulation Low flow rate (large ΔT) Supercritical-pressure ⇒ No phase change No boiling transition No water level Downward-flow water rods ⇒ Large volume fraction Heat sink Large coolant in top dome Super LWR design features water levelcirculation inlet outlet

Safety principle of Super LWR Keeping coolant inventory is not suitable due to no water level and large density change. Coolant inventory is not important due to no circulation. No natural circulation Safety principle is keeping core coolant flow rate. Coolant supply (main coolant flow rate) Coolant outlet (pressure) BWRPWRSuper LWR Requiremen t RPV inventoryPCS inventoryCore flow rate Monitoring RPV water level Pressurizer water level Main coolant flow rate, Pressure

Summary of transient analysis Highest cladding temperature: 730 ℃ = initial temp ℃ Highest pressure: 26.8MPa = 25MPa×1.07 (ABWR: 7.07MPa×1.12) No plastic deformation of cladding Very shot duration of high cladding temp. over 700 ℃ : < 1s, over 670 ℃ : < 5s

Summary of accident analysis “Containment pressure high” signal at Small LOCA (12% cold-leg break): Not considered (like ABWR) ⇒ late ADS (7.6 s) ⇒ PCT= 960 ℃ Considered (like PWR) ⇒ early ADS (1.0 s) ⇒ PCT< 800 ℃ LOCA cladding temperature sensitive to ADS timing (ADS delay)

Summary of ATWS analysis Highest cladding temperature: 890 ℃ (initial temp ℃ ) Highest pressure: 26.8MPa = 25.0MPa × 1.07 (ABWR: 7.07MPa × 1.27) Maximum reactor power: 150% (SCLWR-H) 450% (ABWR)

Start up

Sliding Pressure Startup System Sliding pressure supercritical water-cooled reactor The reactor starts at a subcritical pressure, which increases with load.

Sliding Pressure Startup Curve

Stability Thermal-hydraulic stability Coupled N-TH stability supercritical and sub-critical pressure (start-up)

Plant System HP heaters LP heaters Turbine bypass valve Turbine control valve Exit valve Feedwater pump RPV CR drive Downcomer Water rod Upper dome Lower dome Fuel channel Active core Upper plenum

Stability Criteria The same stability criteria for BWR are used for Super LWR. Normal operating conditionsAll operating conditions Thermal-hydraulic stability Decay ratio  0.5 (damping ratio  0.11) Decay ratio < 1.0 (damping ratio > 0) Coupled neutronic thermal-hydraulic stability Decay ratio  0.25 (damping ratio  0.22) Decay ratio < 1.0 (damping ratio > 0)

Sliding pressure startup curve (Thermal criteria only) Sliding pressure startup curve (Both Thermal and Stability criteria)

Economic potential

Comparison of containments

Super LWRABWRimprovement in % Thermal efficiency, % % RPV weight, t % CV volume, m % Steam line number2450% Turbine speed, rpm3000*1500*50% Condenser2333% *3600rpm and 1800rpm in the western Japan Improvement of 1700MWe Super LWR from 1350MWe ABWR

Super fast reactor (FR) Once-through coolant cycle is compatible with tight lattice fast core because of high head pump and low core flow rate. (Less design limitation than high conversion LWR) Super LWR can be a fast reactor with the same plant system.

Research and Development

Status and Collaboration and Related Researches 1989: Started at The University of Tokyo ; Conceptual design 1998: JSPS-Monbusho funding (University of Tokyo) ; Pulse radiolysis and heat transfer 1998: US DOE-NERI funding (ANL); water-chemistry (pulse-radiolysis) experiments 2000: HPLWR program by EC (FZK, CEA, Framatome, VTT, PSI, KFKI and Univ. of Tokyo) 2001:Japanese NERI of METI funding (Toshiba, Hitachi, Univ. of Tokyo, Kyushu Univ. Hokkaido Univ.) Thermal hydraulics, materials screening, plant concept 2001:US DOE-NERI funding, 3 programs; (INEEL, WH, Univ. Michigan, Univ. of Tokyo, Univ. of Wisconsin, GNF, SRI International, VTT) Corrosion, thermal hydraulics, design studies 2002: Japanese NERI of MEXT (Univ. of Tokyo, CRIEPI, JAERI, Toshiba, Hitachi) Water Chemistry 2002: US Generation 4 reactor (only one among water cooled reactors) 2004: I-NERI between Japan and USA material/corrosion 2005: Japanese NERI of MEXT (Univ. of Tokyo, JAEA, Kyushu Univ. TEPCO) Super Fast Reactor 2006: HPLWR 2 nd phase (Europe), Heat transfer (KAERI)

References of Univ. Tokyo studies SCR2000 symposium, Univ. Tokyo,(2000) first symposium of SCR J.Nucl. Sci.Technol. pp.1081-,(2001),AESJ summary of design ICAPP’02; review of past concepts GENES4/ANP2003; 8 papers; overview, fuel/core, safety, LOCA, control, start-up, stability Global’03; 4 papers; overview, core, ATWS and stability ICAPP’03, ICAPP’05 papers J. Nucl.Sci. Technol. papers