Hee Cheon NO

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Chapter 4 Core Thermal Design and Analysis
Hee Cheon NO Nuclear System/Hydrogen Lab. KAIST
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Contents
4.1 Technical Specifications and Setpoint
Methodology 4.1.1 Standard review plan and
Regulatory guides 4.1.2 Best-estimate methodology
statistical tolerance limits 4.1.3 Technical
Specification and Setpoint Methodology 4.1.4
Software package for Technical Specification
analysis 4.2 Core Thermal Design 4.2.1 Power
peaking factor 4.2.2 Power distribution
control 4.2.3 Fuel centerline temperature
limit 4.2.4 DNB limits Improved DNBR Analysis
Methodology 4.2.5 Practice of PWR core thermal
design 4.3 Core Thermal Analysis 4.3.1 Overview
of subchannel analysis 4.3.2 Subchannel Analysis
Fluid Model of COBRA -?C
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4.1 Core thermal design regulatory guides

• Reactor design The reactor core and associated
  coolant, control, and protection systems shall be
  designed with appropriate margins to assure that
  specified acceptable fuel design limits (SAFDL)
  are not exceeded during and condition of normal
  conditions, including the effects of anticipated
  operational occurrences(AOO).
• Fuel system Design A fuel failure criterion
  should be given for each known failure mechanism.
  Such criteria should address as follows
• Overheating DNB or dryout, fuel melting,
  hydrodynamic flow instability(premature boiling
  crisis)

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2. Reactor design regulatory guides

• Thermal Hydraulic Design DNB correlation
  should be established such that there should be
  95 probability at a 95 confidence level that
  the hot rod in the core does not experience DNB
  for normal operation conditions and AOO. There
  should be at least 99.9 of the fuel rods in the
  core which would not be expected to experience
  departure from nucleate boiling or boiling
  transition for normal operating conditions
• Ex W-3 correlation Limiting DNBR1.3

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2. Reactor design regulatory guides
 
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2. Reactor design regulatory guides
 
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2. Reactor design regulatory guides

• Example of DNBR application of 95/95 concept

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CHF evaluation
Example of CHF evaluation  
   
   
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4.1 Core thermal design regulatory guides

• Technical specification The reactor core and
  associated coolant, control, and protection
  systems shall be designed with appropriate
  margins to assure that specified acceptable fuel
  design limits (SAFDL) are not exceeded during and
  condition of normal conditions, including the
  effects of anticipated operational
  occurrences(AOO).

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2. Reactor design regulatory guides

• Safety limits

      PWR core safety limits
Damage of fuel rod Damage of fuel rod Damage of fuel rod uniform strain of cladding lt 1, MDNBRltlimit DNBR
Accidents LOCA peak cladding temperature lt1204C(2200F)
Accidents LOCA local maximum oxidation of cladding lt17 LOCA cladding embrittlement
Accidents Transients (cladding overheating) maximum fuel centerline temperature No incipient melt
Accidents Transients (cladding overheating) MDNBR MDNBR gt design MDNBR limit
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2. Reactor design regulatory guides

• Safety limits
• Shutdown margin (SDM) margin to criticality in
  the situation with all control rods inserted and
  the strongest control rod withdrawn. SDM is
  initial subcriticality for the steam line break
  accident analysis. sufficient boron concentration
  to assure shutdown without control rod movement
• Fuel enrichment current enrichment limits
  around 5 wt U235 Neither benchmarks of code
  performance nor the bases for extrapolating code
  performance in the enrichment
• range of 5-10 wt have been well established.
• Fuel melting the transition from the solid to
  the liquid phase of UO2 is accompanied by an
  increase (13) in volume neither allow molten
  fuel to contact the cladding nor produce local
  hot spotslt82 kW/m.
• RIA cladding failure maximum radially averaged
  fuel enthalpy of 280 cal/g and DNB failures from
  PCMI and CHF

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2. Reactor design regulatory guides

• Technical Specification Setpoint Development

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2. Reactor design regulatory guides

• Technical Specification Setpoint Development

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2. Reactor design regulatory guides

• Technical Specification Setpoint Development

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2. Reactor design regulatory guides

• Technical Specification Setpoint Development

LSSS
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2. Reactor design regulatory guides

• Technical Specification Setpoint Development

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2. Reactor design regulatory guides

• Technical Specification Setpoint Development

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2. Reactor design regulatory guides

• Technical Specification
• Setpoint Development

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2. Reactor design regulatory guides

• Technical Specification Setpoint Development
• Core Analysis Code Package

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3. Core hydraulic design analysis

• Unit cell concept and ttoal pressure drop with
  spacer grids

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3. Core hydraulic design analysis

• total pressure drop in the core channel
• pressure drop in spacer grid and at inlet and
  outlet
• frictional pressure drop

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3. Core hydraulic design analysis

• pressure drop in spacer grid Rehme's correlation
  

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4. Fuel thermal design analysis

• Fuel centline temperature limit fuel thermal
  analysis methodology
• Issue of the fuel melting the molten fuel may
  contact the cladding leading to cladding
  overheating.
• Fuel centline temperature limit during condition
  II events, the maximum fuel centerline
  temperature should not exceed the fuel incipient
  melting temperature.
• The fuel centerline temperature directly depends
  on the local power generation rate, q(z) with
  the small effects of gap conductivity and melting
  temperature depending on burnup.
• The local power generation rate with the
  incipient melting temperature at the fuel center
  ranges from 18 to 21kw/ft.

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5. Core thermal design analysis

• power peaking factor why do we need it?

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5. Core thermal design analysis

• power peaking factor

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5. Core thermal design analysis

• Example of power peaking factor for rated power
  of 3411Mwth from secondary side heat balance

Energy from fuel 3411Mwth0.9743322Mwth of
active feet of fuel rods (considering
densification, swelling, thermal expansion)
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5. Core thermal design analysis

• estimation of power peaking factor

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5. Core thermal design analysis

• Envelope of limiting peak local power

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5. Core thermal design analysis

• enthalpy rise hot channel factor

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5. Core thermal design analysis

• enthalpy rise hot channel factor

burnup
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5. Core thermal design analysis
CHF concept

 
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5. Core thermal design analysis
CHF correlation(W-3)

 
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5. Core thermal design analysis
MDNBR concept

 
MDNBR design limit1.3
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5. Core thermal design analysis
Core thermal margin Analysis Procedure    
MDNBR failure limit1.0
MDNBR design limit1.3
   
   
MDNBR trend after Rx pump trip
Rx trip time
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5. Core thermal design analysis
DNBR and MDNBR evaluation
   
   
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5. Core thermal design analysis
Example of CHF evaluation  
   
   
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5. Core thermal design analysis
CHF evaluation(modified Biasi correlation)  
   
   
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5. Core thermal design analysis

• Determination of limit DNBR

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5. Core thermal design analysis

• Example of limit DNBRs for DNB correlations

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5. Core thermal design analysis

• Determination of limit DNBR nominal at 95
  confidence level

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5. Core thermal design analysis

• Determination of limit DNBR nominal
• at 95 confidence level

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5. Core thermal design analysis

• Determination of limit DNBR nominal
• at 95 confidence level

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5. Core thermal design analysis

• Determination of norminal
• values uncertainty of
• parameters

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5. Core thermal design analysis

• Example of two DNBR design methods for Loss of
  flow transient

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5. Core thermal design analysis
Core thermal design procedure  
 
   
   
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5. Core thermal design analysis
Example of enginnering Fq
   
   
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- Based on the subchannel control volume where
fluid channels and fuel rods with the
interconnection of neighbor channels - Through
the interconnection of neighbor channels thermal
and momentum mixings by turbulent-induced and
pressure-induced crossflows are considered -
Slip-equilibrium model- The size of each
subchanne a single fluid channel to several
assemblies. - The fuel rod heat transfer model
is coupled with the fluid subchannel analysis
method. - one buffer zone surrounding the hot
channels is accurate enough to the hot
channelsNumerics - fully implicit scheme no
stability limitation on space or time steps-
grid-based marching scheme computation starts
from the inlet grid to the top grid-plane by
grid-plane so that the pressure on each grid
plane may be uniform allowing pressure
equalization through crossflows - valid where
the axial upward flow is dominant.
 
6. Subchannel Analysis
Overview of subchannel analysis
   
   
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6. Subchannel Analysis
Overview
   
   
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6. Subchannel Analysis
Assumptions of subchannel analysis
   
   
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COBRA seriesTORC CELYNX BWVIPRE
EPRINote THINC(W) porous-body code
 
6. Subchannel Analysis
Codes of subchannel anlysis
   
   
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6. Subchannel Analysis
Continuity equation
   
   
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6. Subchannel Analysis
Energy equation
   
   
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6. Subchannel Analysis
Energy equation
   
   
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6. Subchannel Analysis
Energy equation
   
   
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6. Subchannel Analysis
Axial momentum
   
   
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6. Subchannel Analysis
Axial momentum
   
   
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6. Subchannel Analysis
transverse momentum equation
   
   
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6. Subchannel Analysis
transverse momentum equation