ML050400614

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Transcript of ACRS Thermal Hydraulic Phenomena Subcommittee Meeting on Waterford 3 Extended Power Uprate Project, January 26, 2005
ML050400614
Person / Time
Site: Waterford Entergy icon.png
Issue date: 01/26/2005
From:
Entergy Operations
To:
Office of Nuclear Reactor Regulation
References
NRC-194 ACRST-3302
Download: ML050400614 (177)
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Text

Waterford 3 Extended Power Uprate ACRS Thermal Hydraulic Phenomena Subcommittee January 26, 2005

- Entffgy

C.

-z=-En tergy C

Waterford 3 Extended Power Uprate Project ACRS Thermal Hydraulic Phenomena Subcommittee January 26, 2005 I

C-,

A-En tergy Tim Mitchell Engineering Director 2

AdEn tergy Agenda

  • Introduction - Tim Mitchell
  • Safety Analysis - Paul Sicard
  • Risk Considerations - Jerry Holman
  • Engineering Plant Impacts - David Viener
  • Operations Impacts

- Training and Procedures - Gene Wemett

- Testing - David Constance

  • Conclusion - Tim Mitchell 3

C (7

C-Eni ergy Introduction

  • Project Scope
  • Design Basis Improvements
  • Oversight & Rigor
  • Industry Operating Experience 4

(C-

-Entergy Introduction

  • Combustion Engineering Nuclear Steam Supply System (NSSS) Pressurized Water Reactor (PWR)
  • Entered commercial operation 1985
  • 3390 MWt original licensed power
  • 3441 MWt Appendix K Margin Recovery

Ad== En tergy Introduction

  • Project Team
  • Entergy
  • Enercon (Balance of Plant (BOP))
  • Siemens-Westinghouse (Turbine / Generator) 6
=Entergy C-Safety Analysis Paul Sicard EPU Lead Safety Analysis Engineer

(-

7

Entergy Scope of Safety Analysis

  • Demonstrate Acceptable EPU Impact
  • Fuel
  • Non-LOCA Events
  • Containment
  • Radiological 8

Entergy Modification Impact Existing safety systems support safety analyses

  • Replace HP turbine steam path
  • Main Generator rewind
  • Replace Main Generator output breakers
  • Control systems & instrumentation 9

C C-Operating Parameters Parameter EPU Value Current Value Reactor power 3716 MWt 3441 MWt Hot Leg temp 601 OF 600.2 OF Cold Leg temp 541-543 OF 545 OF RCS pressure 2250 psia 2250 psia SG pressure 810 psia 831 psia Steam flow 2301 Ibm/sec/SG 2118 Ibm/sec/SG Feedwater temp 449.7 OF 442.7 OF 10

(7 En tergy Significant Aspects

  • Maintain approximate current nominal Thot
  • Credit ADVs for secondary pressure control for SBLOCA
  • AST methodology for dose calculations 11

C -

C-C-

En tergy Technical Specification Changes Technical Specification changes include:

  • Added ADV Technical Specification
  • Raised minimum BAMT concentration
  • Lowered maximum SIT volume
  • Lowered SG Pressure - Low PPS setpoint
  • Add minimum containment temperature

(-

-'Entergy C-Analysis Changes Parameter EPU Current RCS Cold Leg Temperature Range 536 - 5490F 541 - 5580F (TS 3.2.6)

TCOld Program 541 - 5430 F 5450F ramp constant Minimum Pressurizer Pressure (TS 3.2.8) 2125 psia 2025 psia

  1. SG Tube Plugging Limit 1000 per SG 700 per SG Minimum Boric Acid Makeup Tank (BAMT) 4900 ppm 3950 ppm Boron Concentration (TS 3.1.2.7 and TS Figure 3.1-1) (minor volume changes) 13

-=-En tergy

(

Analysis Changes Parameter EPU Current SG Low Pressure Setpoint (TS Tables 666 psia 764 psia 2-1 and 3.3-4)_

Non-LOCA Transient Analysis Code CENTS CESEC LBLOCA Evaluation Model (EM) 1999 EM 1985 EM Safety Injection Tank (SIT) Level 77.8%

83.8%

Maximum Level (TS 3.5.1) lower plenum lower plenum Post-LOCA Long-Term Cooling (LTC) not credited credited in Approach Changes in mixing mixing volume volume 14

C-Entergy Analysis Changes Parameter EPU Current Fuel Failure for Return to Power Main Yes No Steam line Break (MSLB)

Statistical Convolution for Fuel Failure Yes Yes for selected events Reactor Coolant Radioisotopic ANSI N18.1 ANSI N237 Concentration 15

C-Awb i-En tergy Analysis Changes: Dose Parameter EPU Current Source Term Methodology RG 1.183 RG 1.4 (AST)

(TID-14844)

Primary-to-Secondary Leak Rate (per SG) 75 gal/day 720 gal/day (TS 3.4.5.2)

(operational)

Atmospheric Dispersion Factors New Original license ICRP30 Dose Conversion Factors Yes Yes, for selected events Control Room Doses analyzed (AST)

Only Yes, including LBLOCA and SBLOCA FHA 16

C-

==-En tergy C-Fuel

(-

  • Cycle 14
  • Fuel Mechanical Design Unchanged
  • Standard 16x1 6 fuel design
  • 18 month fuel cycle
  • Erbia burnable poison (since Cycle 9)
  • 217 total assemblies
  • 1 00 fresh assemblies (larger batch size)
  • Acceptable fuel rod corrosion and duty 17

-t

(-

Containment Analysis

& Energy releases releases generated for

  • Peak pressures:

35.16 psig LOCA 41.88 psig MSLB (44 psig acceptance limit) 18

-Entergy Transient Analysis Topics

  • Use of CENTS vs. CESEC:

- CENTS to replace CESEC for non-LOCA transient analyses

- CENTS generically approved for CE plants

  • Credit 3 sec time delay for LOOP after trip for SGTR
  • Demonstrate compliance with acceptance criteria 19

-Entergy Pressurization Events

  • Limiting Fault: Feedwater Line Break 2753 psia (3000 psia acceptance criteria) 20

(7 C-Entergy ECCS Performance Analysis LBLOCA:

  • Update method to 1999 EM (CENPD-132, Supplement 4-P-A)
  • Currently 1985 EM (Supplement 3-P-A)
  • Max Peak Clad Temperature (PCT) of 21640F 21

(

C6-n En tergy ECCS Performance Analysis SBLOCA:

  • No methods change: CENPD-137-P, Supplement 2-P-A (S2M Evaluation Model)
  • Credit automatic operation of ADVs for secondary pressure control

--ADVs Safety Related

  • 1040 psia analysis setpoint
  • Charging Pumps no longer credited
  • 0.055 ft2 break: Max PCT 20180F 22

(7C

-I En tergy ECCS Performance Analysis LOCA Long Term Cooling:

  • Post-LOCA boric acid precipitation analysis assumes mixing volume of core and part of outlet plenum
  • Analysis per CENPD-254 methodology
  • Hot leg injection 2-3 hours post-LOCA demonstrates margin to solubility limit 23

Entergy C-(-

AST Dose Analyses

  • Tracer gas test conducted April 2004
  • License amendment under staff review
  • Bound control room inleakage:
  • Recirculation Mode:

100 CFM (79 CFM measured)

  • Pressurized Mode: 65 CFM (36 CFM measured) 24

i Entergy AST Dose Analyses

  • Analyses extended to non-LOCA radiological events and Small Break LOCA
  • High Control Room X/Q due to proximity of ADVs to Control Room Air Intakes Assume leakage of 0.375 GPM for faulted SG (MSLB, FWLB)
  • Assume 150 gal/day for intact SGs (75 gal/day TS limit only for SBLOCA)
  • Credit existing operator action to select preferred control room air intake 25

C-C-

C7 AST Results Results for Limiting Events:

Fuel EAB LPZ MCR Failure TEDE TEDE TEDE I.C. MSLB 10%

0.60 0.19 4.89 FWLB / O.C. MSLB 0%

0.23 0.12 3.62 CEA Ejection 15%

1.03 0.65 2.41 SGTR (PIS) 0%

0.99 0.21 4.85 LBLOCA RG 1.183 5.30 2.37 2.95 SBLOCA 100%

1.96 1.08 3.93 FHA 60 rods 0.55 0.085 0.11 26

(-7

-=- En tergy AST Dose Analyses Conclusions

C a=-

En tergy

.f Risk Considerations Jerry Holman Manager, Nuclear Engineering 28

Aftr AdEnt'Cry Scope Of Risk Assessment Address Impact On

  • Initiating Event frequency
  • Success criteria
  • Equipment failure rates
  • Operator response times and Human Reliability Analysis (HRA)

CDF and LERF

  • External events
  • Shutdown 29

C

=Entergy C7-Risk Assessment Results

  • Initiating Event Frequency
  • No new initiators
  • No change in frequencies
  • Success Criteria
  • CENTS analyses to confirm success criteria
  • No changes 30

C (7c-

-Entergy Risk Assessment Results Equipment Failure Rates

  • Comprehensive reviews of equipment performed
  • Systems operate within allowable limits
  • No impact on PRA failure rates or results
  • Existing monitoring programs and model update will account for any additional system wear 31

6-MEnterggy Q71 Risk Assessment Results

  • Operator Response Times / HRA
  • CENTS analyses to determine available action times
  • Higher decay heat reduced operator action times
  • Major impact is reduction of recovery of feedwater time for loss 32

C-

`En tegy C-Risk Assessment Results Scenario Pre-EPU Time Post-EPU Available Time Available Recover feedwater for early loss of 82.6 min 68.3 min feedwater Recover feedwater for late loss of 5.1 hr 4.1 hr feedwater (battery depletion)

Recover feedwater for late loss of 12.3 hr 11.3 hr feedwater (CSP depletion) 33

AM C, (7

(is /

Entergy Risk Assessment Results

  • Internal Events (per year):
  • CDF increase = 3.5E-7
  • LERF increase < 1.OE-7
  • New CDF = 5.9E-6 34

=-- En-lergy Risk Assessment Results EPU Sequence Contribution Total Loss of Feedwater 45%

Other 27%

Station Blackout 28%

35

C.

C-

=-- En tergy Risk Assessment Results

  • External Events
  • Slight increase in fire CDF due to operator response time reduction
  • No impact on other external events 36

(

(

En tergy Risk Assessment Results
  • Shutdown Risk
  • EPU has no unique or significant impacts
  • No changes to shutdown operations protection plan 37

C Cc

-~Entergy Risk Assessment Results Conclusions

  • All PRA model elements reviewed for impact
  • Minor reduction in Operator recovery times
  • EPU has a very small impact on risk 38

C-Z-Entergy C(

Engineering Plant Impacts David Viener EPU Lead Mechanical Engineer 39

aM En terU Significant Modifications

  • Replace HP turbine steam path
  • Main Generator rewind & alkalizer skid
  • Replace Main Generator output breakers

I ~En t ergy C-Significant Modifications (cont.)

  • FW heater drain valve capacity increase
  • Condenser tube staking
  • Control systems & instrumentation
  • Setpoint, range and scale changes
  • 4 transmitters to be replaced 41

AdEntergy Engineering Plant Impacts Decay Heat Ultimate Heat Sink

  • System Capable of Dissipating Heat Loads for Normal, Shutdown and Accident Conditions
  • Water Sources are Adequate to Maintain Cooling of Essential Plant Equipment Equipment Operating Times Increased Post-Accident which Impacts Emergency Generator Fuel Oil 42

C C-C A-En tergy Engineering Plant Impacts

  • Decay Heat
  • Raised fuel oil minimum capacity requirement to maintain 7 day supply per current licensing basis.
  • Commitment to add additional storage.

43

Entergy Engineering Plant Impacts

  • Decay Heat (Cont'd)
  • System Flow Capable of Mitigating against Feedwater Demand Events
  • Normal and Backup Condensate Sources are Adequate to Bring Plant to Shutdown Cooling Entry Conditions 44

(I-(z-

-~En tergy Engineering Plant Impacts

  • Decay Heat (Cont'd)

Shutdown Cooling Capable of Achieving Cold Shutdown Conditions in accordance with Reactor Systems Branch (RSB) Branch Technical Position (BTP) 5-1

  • Refueling Technical Specification Time Limits to Reduce Shutdown Cooling Flow remain Unchanged 45

(

(c-

-Entergy Engineering Plant Impacts Decay Heat (Cont'd)

Fuel Pool Cooling

  • EPU Proposes a 1.5% Increase
  • Current Fuel Pool Temperature Limits will be Maintained

C

(

C..

Entergy Engineering Plant Impacts

  • Containment Overpressure
  • Containment Overpressure not Credited in the ECCS Pumps Net Positive Suction Head Analysis
  • EPU Maintains this Assumption
  • Systems Inside Containment will be Unchanged
  • Minimum Containment Water Level remains Unchanged
  • Sump Temperature change is Negligible 47

En tegy Engineering Plant Impacts Vibration

  • Detailed tube bundle evaluation
  • Dryers and Dryer Supports evaluated
  • Palo Verde Dryer Design - Operating at Higher Flow Rates than W3 Proposes.
  • Secondary System
  • Condenser Tube Staking Required
  • Vibration Monitoring Program
  • Monitor Secondary Systems pre-and post-EPU based on Industry Operating Experience.

48

C C

En htergy Engineering Plant Impacts

  • No component replacements required
  • Outage inspection sampling increased based on EPU conditions
  • Piping systems impacted will continue to be monitored to detect any deviation from predicted wear rates.

49

C

(

C

'Entergy Engineering Plant Impacts

  • Alloy 600

- Nominal Thot increasing by 0.8 OF

  • Nominal Tcold decreasing by 2 OF
  • Impact on crack initiation rate is negligible

C C'

-a-Enterg~y Engineering Plant Impacts

  • Grid Stability
  • Short Circuit, Transient Stability and Offsite Voltage Stability Studies Re-performed
  • Short Circuit Study Determined Generator Output Breakers were marginal
  • Installing larger generator output breakers for EPU 51

A(

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A-Entergy Engineering Plant Impacts Conclusion With the proposed modifications, Waterford 3 plant design can safely operate at the proposed EPU conditions 52

'Entergy C

Operator Impact/Training Gene Wemett Assistant Operations Manager C

53

(

C-C En tergy Operator Impact/Training

  • Operations oversight
  • Review of all modifications and evaluations for impact on operation
  • Procedure impact 54

C~

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Q OEntergy Operator Impact/Training Training

  • Phase 1, EPU seminars on modifications, Technical Specification (TS) changes and procedure changes (complete)
  • Phase 11, Crew training on plant modifications (in progress)
  • Phase Ill, Crew training on procedure changes, setpoint changes, TS changes (begins in March)
  • Crews evaluated on the uprated plant simulator prior to refueling outage
  • Crews evaluated on TS, procedure and setpoint changes 55

--Entergy Operator Impact/Training Controls and Displays

  • Changes minimal
  • Change to allow more precise setting of Atmospheric Dump Valve setpoint
  • Turbine will be operated exclusively in single valve
  • Some display ranges will be re-scaled 56

(

(

(

En tegy Operator Impact/Training

  • Technical Specifications (TS)
  • Parameter changes
  • One new Atmospheric Dump Valve TS
  • Normal and Off-normal Procedures
  • No new procedures
  • Emergency Operating Procedures
  • No change to type and nature of actions
  • No new actions 57

(

I C

C

- En tergy Operator Impact/Training Conclusion The changes brought about by power uprate to unit's operation are minimal and acceptable to the Operations Department.

58

C Power Ascension Testing David Constance Operations C

59

c E~~En tergy Power Ascension Testing (IS.

  • Reactor Engineering Tests / Power Verification
  • Post Modification Testing
  • Plant Maneuver Test (1 00%-90%-95%)
  • Post 100% Testing, Data Collection &

Surveys

  • Vibration Monitoring 60

Ad Entergy Power Ascension Profile 100.0%

90.0%

80.0%

70.0%

60.0%

50.0%

40.0%

30.0%

20.0%

10.0%

95%

9 100/

92.5%

30 hrs 68%

15 hrs 50%

-10 0

10 20 30 40 50 60 70 80 90 100 110 120 130 Hours 140 150 61

(

C Ad En terg Power Ascension Testing

  • Low Power Physics Testing (LPPT) remains unchanged for EPU
  • Data sets
  • Collected every 10% from 20-100%
  • Collected at 7 different power -plateaus
  • Approximately 1000 parameters monitored
  • Data will be automatically collected and processed
  • Data evaluated against predetermined criteria
  • Plant Safety Subcommittee reviews results report at each power plateau (>68%), and recommends continued power ascension.

62

(

(

C Entergy Testing Considerations

  • The proposed plant modifications either have
  • No significant impact on transient response, or
  • Have been evaluated using a calculation model l No physical changes to the Nuclear Steam Supply System
  • No new interactions that affect system response
  • No changes to controller algorithms 63

Entergy Testing Considerations

  • Post Modification Testing demonstrates that the component/systems will perform as designed
  • Power ascension data collection confirm acceptable operation
  • Maneuvering test provides further confirmation
  • Benchmarked calculational model evaluates postulated transient conditions 64

Power Ascension Testing Conclusion

  • The planned post modification testing and startup tests confirm that the analyses, modifications and adjustments necessary for EPU have been completed properly
  • Adequate safeguards are in place to insure a controlled, closely monitored, conservative approach to the new licensed power level 65

MEntergy Concluding Remarks Tim Mitchell Engineering Director

(

66

c

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End of Presentation C

67

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'4l iA a I Reactor Systems Branch Audit Calculations L. W. Ward US Nuclear Regulatory Commission Division of Systems Safety and Analysis Reactor Systems Branch ACRS Meeting January 26, 2005

I Reactor Systems Branch Audit Calculations Agenda o

Large Feedwater Line Break o

Limiting Small Break LOCA o

Post-LOCA Long Term Cooling (Boric Acid Precipitation and Timing for Simultaneous Hot/Cold Side Injection)

REACTOR SYSTEM BRANCH AUDIT CALCULATIONS Waterford EPU o

Large Feedwater System Pipe Break

- Alternate Methodology Verified Peak RCS Pressure

- Conservative Analysis Assumptions (break at the elevation of the tube sheet)

Waterford 3 Extended Power Uprate 2900 -

COLD LEG PRESSURE 2750 -

0 - -0 Staff Calculation 2500 ccc M

4020460810 CTM S

wC' 245R PRESSURE PRES Prsuev.Tm 0

2300 2150 2000 2

2 03005 020 40 60 80 100.

TIME, SECONDS Figure 2.13.2.3.1 -2 Feedwater System Pipe Break (Large)

RCS Pressure vs. Time 6306-1.doc-i 1/05/03 2.13-1 88 6306-l.doc-1 1/05/03 2.13-188

Waterford 3 Extended Power Uprate 120 10 0

G -T -

Staff Calculation CD 00 0

0<

6 U

~~~~~SG V2udMssv.Tm LU 4 0 SG

/

20 A.A I* 020460s 0

TIME SEOND 0F20u40 6013823100 Feedwater System Pipe Break (Large)

K....-'SG Liquid Mass vs. Time 8306-1.doc-i 1105/03 2.13-195 6306-l.doc-1 1/05/03 2.1 3-195

Con't o

Limiting Small Break LOCA in the Pump Discharge Leg

- Staff Calculations Reproduced CEFLASH-4AS Core Transient Two-phase Level for the Limiting Small Break(Q.055 ft2 CLB)

- No Credit for Accumulator Injection

- Conservative Analysis Assumptions (Top Skewed Axial shape, Diesel Failure, 1.2 Decay.Heat Multiplier) to W3F1-2004-0052 Page 16 of 32 F1giure 2.12-45 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2/PD Break InnerVessel Pressure 2400 2000 1600 05 EL Ui so Ura:

1200 l

I 0

CG - -0e Staff Calculation In=

800 400 0

)

0 600 1200 1800 2400 3000 TIME, SEC to W3F1-2004-0052 k_

Page 19 of 32 Figure 2.1248 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 f12JPD Break Inner Vessel Two-Phase Mixture Level 48 40 I'

LU zCI 32 24 16 1..........

) - - e Staff Calculation SOTCLO OF COR.I

i., _

1....

8 B

0 600 1200 1800 2400 3000 TIME, SEC

Con't o

Post-LOCA Long Term Cooling (Prevention of Boric Acid Precipitation)

- Staff Calculations Revealed Error in Mixing Volume ( assumed void fraction of 0% in mixing volume following LB LOCAs)

- Error Produces Precipitation at One Hour vs Four Hours

- Westinghouse has Corrected Error and Modified Licensing Methodology Mixing Volume Reflects Liquid in Core and Upper Plenum to Hot Leg Top EL (vs mixing vol to hot leg bottom elevation)

Minimum Containment Pressure Raised to 20 psia (vs 14.7 psia)

Performed Min. Cont. Pressure Calculation using NRC Approved Methodology (GOTHIC)

COLD LEG BREAK I

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Boron Concentration vs. Time Waterford EPU, No Core Flushing Flow 36 30 solubility limit = 27.6 wt%

an18

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0 124 0

./void=0.0, mixing vol core +up below h 6...../..-

void=0.63, mixing, vol core~up below hl L/

0 0

2 4

6 8

10 Time, hr

Boric Acid Concentration vs. Time C

Waterford EPU, Effect of, Containment Pressure 40 0'

0' 36/

0 32 0C 20psialimit 28 :

,14.7 psia limit 24

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o cis)2 16 12 o -- e Containment Pressure 14.7 psia 8

l3 - - El Containment Pressure 20 psia

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4 6

Time, hrs

Con't

- Staff Believes Adequate Margin Remains to Support Power Uprate No Credit for Liquid Entrainment (also no removal of boric acid by vapor)

No Mixing in hot Legs Boric Acid Make-Up Tanks, BAMTs, Discharge (6187 ppm)

Upper Plenum Pressure Higher than Cont. by Loop Pressure Drop

- Westinghouse will Document Changes to Methodology and Revised Analyses

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