ML15113A787

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CEA Drop Time Technical Specification Change Request April 22, 2015 Revised Public Meeting Slides
ML15113A787
Person / Time
Site: Waterford Entergy icon.png
Issue date: 04/22/2015
From:
Entergy Operations
To: Michael Orenak
Office of Nuclear Reactor Regulation
Wang A
References
Download: ML15113A787 (69)
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Text

CEA DROP TIME TT.S.

S CHANGE REQUEST WATERFORD 3 APRIL 22, 2015 1

Licensee Attendees

  • Waterford 3

- John Jarrell - Manager, Regulatory Assurance

- Pamela Hernandez - Supervisor, Reactor Engineering

- Leia Milster - Licensing Engineer, Regulatory Assurance

- Willi William St l Steelman - Contractor, C t t SMI

  • Westinghouse

- Kim Jones - Fellow Engineer, g Setpoints, p Controls & Containment

- Matthew Wilcox - Senior Engineer, Setpoints, Controls & Containment

- Amanda Maguire - Senior Engineer, Regulatory Compliance 2

Outline

  • Background for the License Amendment Request (LAR)
  • LAR Technical Content

- Control Element Assembly (CEA) SCRAM Insertion Curve position vs. time

- Safety Analysis Margin

- Core Operating Limit Supervisory System (COLSS) Required Overpower Margin (ROPM)

  • Schedule 3

Background for LAR

  • CEA Drop Times have challenged the Technical Specification (TS) limit in the last two surveillance performances

- Waterford 3 TS 3.1.3.4 requires:

  • the arithmetic average of all CEA Drop Times be 3.0 seconds
  • Individual CEA drop times 33.2 2 seconds
  • Insertion time is measured from fully withdrawn position to 90% inserted 4

Historical Drop Times CEA drop time group arithmetic average 3200 3100 Time tto 90% (millliseconds) 3000 2900 2800 2700 2600 2500 2400 Cycle Number

  • Cycle 20 includes repeated test 5

Potential Causes Plant Primary Side Modifications

  • Reactor Vessel Head replacement
  • CEA replacement
  • Transition to Next Generation Fuel Product 6

Proposed TS Change

- Raise the arithmetic average of all CEA Drop Times to be 3.2 seconds

- Raise the Individual CEA drop times to 3.5 seconds 7

Analysis Goal Overall

  • The Chapter 15 Safety Analyses continue to meet all acceptance criteria
  • The current Licensing Basis will remain bounding for the revised analyses
  • Th The are no changes h tto th the COLSS and d CPCS d databases t b andd addressable constants.

8

CEA SCRAM Insertion Curve Average Position vs Time 120 100 80 Percent With hdrawn 60 40 20 0

0 1 2 3 4 Time Safety Analysis SCRAM Curve Safety Analysis SCRAM Curve with Delay 9

Safety Analysis Basis Analysis Margin

  • Safety analysis input (SAIs) consist of plant design and operating parameters that include uncertainties associated with them.
  • Safety analyses apply the uncertainties in a conservative, or more restrictive, t i ti di direction.

ti

  • On some parameters, the SAI uses a bounding input value, which is more adverse than the plant parameter plus uncertainty uncertainty.
  • Thus, the bounding SAI includes analysis margin, which can be reclaimed later.

10

Safety Analysis Margin Revised Reactivity Safety Analysis Input vs Time Time 0.2 0 1 2 3 4 0

-0.2

-0.4 Reactivity

-0.6

-0.8

-1

-1.2 Current Reactivity SAI Current Reactivity SAI with Delay Revised Reactivity SAI 11

Safety Analysis Margin COLSS ROPM

  • Waterford is a Combustion Engineering Design Digital Plant

- Limiting Conditions for Operations (LCOs)

  • Technical Specifications
  • Core Operating Limits Supervisory System

- Maintains Departure from Nuclear Boiling Ration (DNBR)

( ) Margin

>> Required Overpower Margin (ROPM)

>> Linear Heat Rate (LHR)

- DNB and LHR Protection

  • Core Protection Calculator Systems (CPCS) 12

Safety Analysis Margin COLSS ROPM (contd)

  • ROPM

- Event ROPM is the actual thermal margin change during the design basis event (DBE).

- Initial Analysis ROPM is the thermal margin set aside by the Safety Analyses at the start of the event event.

- COLSS ROPM is the thermal margin reserved by the COLSS to support the Technical Specification LCOs.

- If COLSS ROPM > Event ROPM, then the minimum DNBR (mDNBR) > DNB Specified Acceptable Fuel Design g Limit (SAFDL)

( )

- ROPM can be used to offset calculated fuel failures due to DNB.

13

Method of Analysis Analysis Basis

  • There are no changes to the LOCA or Non-LOCA transient analysis methods from those in the current UFSAR.
  • There are no changes to the Core Design / Neutronics methods that provide input to the LOCA and Non-LOCA transient analyses that h support theh current UFSAR UFSAR.

14

DNBR Correlation Analysis Basis

  • There are no changes to the DNBR critical heat flux correlation.
  • The DNB SAFDL value remains unchanged unchanged.

15

Topical Reports Applicability

  • The change in CEA SCRAM Insertion Curve does not impact the topical reports cited by Waterford Unit 3.
  • Non-LOCA case results presented in the topical reports provide illustrative examples to confirm methodology, simulations, and d

determinei trends.

d

  • Conservative selection of inputs are performed in the plant specific analyses to support the UFSAR.

16

UFSAR Chapter 15 Non-LOCA Analyses Event Categorization

  • Each Non-LOCA UFSAR Chapter 15 Events will be discussed
  • For each event event, they are categorized as:

- Evaluated (Impact analysis and/or evaluation utilizing current methodology)

- Assessed A d (I (Impactt determination d t i ti and d jjustification tifi ti provided) id d)

- Bounded by another DBE

- Not Impacted 17

A l Analyses Evaluated E l t d 18

Analyses Evaluated Chapter 15.1.1.3: Increased Main Steam Flow

  • Impacted by revised CEA SCRAM Insertion Curve Curve.

- ROPM increases ~1%.

  • Evaluated

- Benefit of Revised Reactivity Safety Analysis Input

  • ROPM decreases by ~0.6% for initial power 50%.
  • No benefit for initial p power < 50%.

- Hot Full Power (HFP)

  • COLSS HFP ROPM > Initial Analysis ROPM

- Intermediate power levels for CPCS

  • Both trip and no-trip cases are analyzed.
  • No-trip cases bound trip cases.

19

Analyses Evaluated Chapter 15.1.1.3: Increased Main Steam Flow

  • Summary

- Hot Full Power

  • Revised Reactivity SAI and COLSS ROPM offset the revised CEA SCRAM Insertion Curve.

- Intermediate Power

  • No-trip cases are expected to remain bounding.
  • Conclusions

- N No changes h to COLSS M Margini expected.

d

- No changes to CPCS Input expected.

20

Analyses Evaluated Chapter 15.1.2.3: Increased Main Steam Flow with Single Failure (SF)

  • Supports radiological dose fuel failure limit of 8%.
  • Analyzed at HFP HFP.

- Cycle specific fuel failure is ~50% of the limit.

  • Impacted by revised CEA SCRAM Insertion Curve.

- Expected fuel failure to increase by ~2%.

  • Evaluated

- Revised Reactivity SAI to lower expected fuel failure ~1%.

~1%

21

Analyses Evaluated Chapter 15.1.2.3: Increased Main Steam Flow with Single Failure

  • Summary of the Combined Impact

- Expected fuel failure to increase by ~1%.

1%.

  • Conclusions

- Calculated fuel failure is expected to be < 8%.

- No change to the radiological dose results expected.

22

Analyses Evaluated Chapters 15.2.1.3/15.2.2.3: Loss of Condenser Vacuum (LOCV)

(w/wo SF)

  • Impact estimate based on engineering judgment

- Estimated peak RCS pressure increase < 1 psi.

- Estimated peak SG pressure increase < 2 psi.

psi 23

Analyses Evaluated Chapters 15.2.1.3/15.2.2.3: LOCV (w/wo SF)

  • Evaluated

- Impact of revised CEA SCRAM Insertion Curve on peak RCS and SG pressure evaluated to determine increases.

  • < 1 psi prior to 90% insertion for RCS pressure.
  • < 5 psi after 90% insertion for RCS pressure.
  • < 1 psi for SG pressure.

- Benefit of Revised Reactivity SAI on peak RCS and SG pressure evaluated to determine decreases.

  • 0 psi prior to 90% insertion for RCS pressure.
  • > 3 psi after 90% insertion for RCS pressure.
  • 0 psi for SG pressure pressure.

24

Analyses Evaluated Chapters 15.2.1.3/15.2.2.3: LOCV (w/wo SF)

  • Summary of the combined impact

- Prior to 90% insertion - Peak RCS pressure increases < 1 psi.

- After 90% insertion - Peak RCS pressure increases < 2 psi.

- Peak SG pressure < 1 psi.

  • Confirmed Engineering judgment 25

Analyses Evaluated Chapters 15.2.1.3/15.2.2.3: LOCV (w/wo SF)

  • Current Chapter 15 Results

- Current peak RCS pressure = 2711 psia < 2750 psia criterion.

- Current peak SG pressure = 1181 psia < 1210 psia criterion.

  • Conclusions

- Updated peak RCS pressure < 2750 psia criterion.

- Update peak SG pressure < 1210 psia criterion.

  • Results are used to confirm assessments on subsequent DBEs.

26

Analyses Evaluated Chapter 15.3.2.1: Total Loss of Forced Reactor Coolant Flow

  • Impacted by revised CEA SCRAM Insertion CurveCurve.

- ROPM increases ~1%.

  • Evaluated

- Safety analysis margin benefit

  • Revised Reactivity SAI
  • COLSS HFP ROPM > Initial Analysis y Margin g
  • Summary

- No impact on the results and conclusions.

27

Analyses Evaluated Chapters 15.3.3.1/15.3.3.2: Single Reactor Coolant Pump (RCP) Shaft Seizure/Single RCP with a stuck open secondary safety valve

  • Supports pp radiological g dose fuel failure limit of 15%.
  • Analyzed at HFP.

- Cycle specific fuel failure is ~50% of the limit.

  • Impacted by revised CEA SCRAM Insertion Curve.

- Expected fuel failure to increase ~2%.

  • Evaluated

- Benefit of revised Reactivity SAI to lower expected fuel failure ~1%.

28

Analyses Evaluated Chapters 15.3.3.1/15.3.3.2: Single Reactor Coolant Pump (RCP) Shaft Seizure/Single RCP with a stuck open secondary safety valve

  • Summaryy

- Expected fuel failure to increase ~1 %.

- Total cycle specific fuel failure < 15 %.

- Insignificant impact on steam releases.

releases

- No recalculation of radiological doses.

  • Conclusion

- No changes to radiological doses.

- No changes to COLSS ROPM.

29

Analyses Evaluated Chapter 15.4.1.1: Uncontrolled CEA Withdrawal from a Subcritical Condition

  • Supports pp

- mDNBR

- Fuel Melt Limit

  • Current C tRResults lt

- mDNBR >> DNB SAFDL

- Fuel Centerline Temperature << Fuel Melt Limit

  • Impacted by revised CEA SCRAM Insertion Curve.

30

Analyses Evaluated Chapter 15.4.1.1: Uncontrolled CEA Withdrawal from a Subcritical Condition

  • Evaluated/Assessed

- Expectation that mDNBR >> DNB SAFDL.

- Expectation that Fuel Centerline Temperature << Fuel Melt Limit.

  • Conclusions

- Negligible impact on the results and conclusion.

- No change to the UFSAR.

31

Analyses Evaluated Chapter 15.4.1.2: Uncontrolled CEA Withdrawal at Low Power

  • Supports

- mDNBR

- Fuel Melt Limit

  • Current Results

- mDNBR >> DNB SAFDL

- Fuel Centerline Temperature << Fuel Melt Limit

  • Impacted by revised CEA SCRAM Insertion Curve.
  • Evaluated/Assessed E l d/A d

- Expectation that mDNBR >> DNB SAFDL.

- Expectation that Fuel Centerline Temperature << Fuel Melt Limit.

  • Conclusions

- Negligible impact on the results and conclusion.

- No change g to the UFSAR.

32

Analyses Evaluated Chapter 15.4.1.3: Uncontrolled CEA Withdrawal at Power

  • Impacted by revised CEA SCRAM Insertion Curve.

- ROPM increases ~1%.

  • Evaluated

- Benefit of Revised Reactivity Safety Analysis Input

  • ROPM decreases by ~0.6% for initial power 50%.
  • No benefit for initial power < 50%.

- HFP

  • COLSS HFP ROPM > Initial Analysis ROPM

- Intermediate power levels for CPCS

  • Both h trip i andd no-trip i cases are analyzed.

l d

  • No-trip cases bound trip cases.

33

Analyses Evaluated Chapter 15.4.1.3: Uncontrolled CEA Withdrawal at Power

  • Summary

- HFP: Revised Reactivity SAI and COLSS ROPM offset revised CEA SCRAM Insertion Curve.

- Intermediate Power: No-trip cases are expected to remain bounding.

  • Conclusions

- No changes to COLSS Margin expected.

- No changes to CPCS Input expected.

34

Analyses Evaluated Chapter 15.4.3.6: CEA Ejection

  • Supports radiological dose fuel failure limit of 15% for DNB and 0% for fuel rod enthalpy.
  • Analyzed parametric in power from HFP to Hot Zero Power (HZP).

- Cycle specific fuel failure is ~70% of the limit.

  • Defines key Non-LOCA Input.

- COLSS ROPM for HFP and intermediate power levels

- CPCS input

- Bounding physics input

  • Impacted by revised CEA SCRAM Insertion Curve.

- Expected fuel failure to increase by ~2%.

- Expected rod enthalpy to increase above the limit.

- Insignificant g for steam releases.

35

Analyses Evaluated Chapter 15.4.3.6: CEA Ejection

  • Evaluated

- Benefit of revised Reactivity SAI partially offsets the revised CEA SCRAM Insertion Curve for initial power levels 50%.

  • Expected to lower fuel failure ~1%.
  • Expected benefit to offset ~50%

50% of the enthalpy increase increase.

- If calculated fuel failures exceed radiological dose limits, OR if fuel rod enthalpies exceed the limits, THEN:

  • Credit C dit SAI margin i iin b bounding di physics h i d data t ffor allll power llevels l if needed to maintain current results.
  • Reduce bounding physics values for ejected CEA rod worth and ejected peaks.

k 36

Analyses Evaluated Chapter 15.4.3.6: CEA Ejection

  • Summary

- Reduction in bounding ejected CEA rod worth expected.

- Reduction in bounding ejected CEA peak expected.

- No changes to COLSS ROPM expected.

- N changes No h to t CPCS iinputt expected.

t d

  • Conclusion

- Expected p impact p on fuel failure to remain < 15% for DNB and 0% for fuel rod enthalpy.

37

Analyses Evaluated Chapter 15.9: Asymmetric Steam Generator Transient

  • Supports COLSS ROPM and CPCS input.
  • Impacted by revised CEA SCRAM Insertion Curve Curve.

- ROPM increases ~1%.

  • Evaluated

- Benefit of Revised Reactivity Safety Analysis Input

  • ROPM decrease by ~0.6% for initial power 50%.
  • Initial Analysis ROPM > Event ROPM
  • COLSS ROPM > Initial Analysis ROPM 38

Analyses Evaluated Chapter 15.9: Asymmetric Steam Generator Transient

  • Summary

- HFP

  • Revised Reactivity SAI and COLSS ROPM offset the impact of revised CEA SCRAM Insertion Curve.

- Intermediate Power

  • Revised Reactivity SAI and Initial Analysis ROPM offset the impact of revised CEA SCRAM Insertion Curve.
  • Conclusion C l i

- No change to the COLSS Margin expected.

- No change to the CPCS Input expected.

39

A l Analyses Assessed A d 40

Analyses Assessed Chapter 15.1.1.4: Inadvertent Opening of an Atmospheric Dump Valve

  • Hot Zero Power

- mDNBR

- LHR

- Steam releases

  • Current C R Resultsl

- mDNBR >> DNB SAFDL at ~85 seconds

- LHR << Steadyy State Limit at ~83 second

- Reactor trip occurs at 600 seconds hour steam releases = ~1 M-lbm

- Shutdown cooling steam releases = ~2 2.5 5 M-lbm M lbm 41

Analyses Assessed Chapter 15.1.1.4: Inadvertent Opening of an Atmospheric Dump Valve

  • Impact of the revised CEA SCRAM Insertion Curve

- No Impact on mDNBR and peak LHR.

- Insignificant impact on steam releases.

  • Conclusion

- No impact on the results and conclusions.

42

Analyses Assessed Chapter 15.1.2.4: Inadvertent Opening of an Atmospheric Dump Valve with SF

  • Hot Full Power

- Supports radiological dose; fuel failure limit is zero.

  • Impacted by revised CEA SCRAM Insertion Curve

- mDNBR < DNB SAFDL

  • Assessed/Evaluated

- Benefit of revised Reactivity SAI to increase mDNBR

- COLSS HFP ROPM > Initial Analysis ROPM 43

Analyses Assessed Chapter 15.1.2.4: Inadvertent Opening of an Atmospheric Dump Valve with SF

  • Impact of CEA SCRAM Insertion Curve is offset by benefit of safety analysis margin.

- Revised Reactivity SAI

- Analysis Margin in COLSS HFP ROPM value

- mDNBR > DNB SAFDL

- No Fuel Failure

  • Conclusions

- No impact on the results and conclusions.

44

Analyses Assessed Chapter 15.1.3.1: Steam System Piping Failures Post-trip Return-to-Power (R-t-P) and Return-to-Criticality (R-t-C)

  • Hot Full Power and Hot Zero Power w/wo LOAC
  • Assessed

- Insignificant impact on steam releases.

- Rate of reactivity insertion during the CEA SCRAM rod insertion has a negligible impact on the reactivity balance at the time of R-t-P and R-t-C.

  • Conclusions

- No impact on the results and conclusions.

45

Analyses Assessed Chapter 15.1.3.3: Steam System Piping Failures Pre-trip Power Excursion Analysis

  • Supports radiological dose fuel failure limit of 8%.

- Current calculated fuel failure is zero.

- mDNBR >> DNB SAFDL

  • Impacted by revised CEA SCRAM Insertion Curve.
  • Assessed

- Use results from the Increased Main Steam Flow

- COLSS ROPM > Event ROPM 46

Analyses Assessed Chapters 15.1.3.3: Steam System Piping Failures Pre-trip Power Excursion Analysis

  • Impact p of the CEA SCRAM Insertion Time offset byy safetyy analysis y

margin.

- Revised Reactivity SAI

- COLSS HFP ROPM > Event ROPM

- mDNBR > > DNB SAFDL

- Calculated fuel failure to remain zero.

  • Conclusions

- No impact on the results and conclusions.

47

Analyses Assessed Chapters 15.2.2.5/15.2.3.2: Loss of Normal Feedwater Flow (w/wo SF)

  • Assessed using LOCV results.
  • Expectation is the benefit of the Revised Reactivity SAI offsets the revised CEA SCRAM Insertion Curve.

48

Analyses Assessed Chapter 15.2.3.1: Feedwater System Pipe Breaks

  • Assessed using LOCV results.
  • Expectation is the benefit of the Revised Reactivity SAI offsets the revised CEA SCRAM Insertion Curve.

49

Analyses Assessed Chapter 15.4.1.4: CEA Misoperation: Single CEA Withdrawal (SCEAW)

  • Impacted by revised CEA SCRAM Insertion Curve.
  • Current results

- Analysis performed at intermediate power levels.

- Both trip and no-trip cases are analyzed.

- No-trip cases bound trip cases.

  • Assessed using CEAW at power results.results
  • Impact

- Benefit of revised Reactivity SAI is available to offset some of the increase for p cases due to revised CEA SCRAM Insertion Curve for initial p the trip power levels 50%.

- No-trip cases are expected to bound the trip cases.

  • Conclusions

- No change h to COLSS Margin expected.

- No change to CPCS Input expected.

- No changes to results and conclusions are expected.

50

Analyses Assessed Chapter 15.6.3.2: Steam Generator Tube Rupture

  • Supports radiological doses.

- Primary Primary-to-secondary to secondary mass transfer

- Steam releases

  • Assessed

- Rate of reactivity insertion during the CEA SCRAM has an insignificant impact on the primary-to-secondary mass transfer and secondary steam releases.

  • Conclusions

- No impact on the results and conclusions expected.

51

Analyses Assessed Chapter 15.6.3.3: LOCA

- SCRAM rod insertion not credited.

- Not Impacted

- The expectation is that the impact will be negligible.

  • Long term cooling

- SCRAM rod insertion is not credited credited.

- Not Impacted

  • Conclusions

- No changes to the results and conclusions are expected.

52

Analyses Assessed Chapter 15.8: Anticipated Transient Without SCRAM

  • Diversified SCRAM System setpoints not impacted
  • Rate of reactivity insertion during the CEA SCRAM has a negligible impact on the results.
  • Not Impacted 53

Analyses A l Bounded B d d by b Another A th Analysis y

54

Analyses Bounded by Another Analysis

  • Chapters 15.1.1.1/15.1.2.1: Decrease in Feedwater Temperature (w/wo SF)

- Bounded by the Increased Main Steam Flow (w/wo SF) in Chapters 15.1.1.3/15.1.2.3.

  • Chapters 15.1.1.2/15.1.2.2: Increase in Feedwater Flow (w/wo SF)

- Bounded by the Increased Main Steam Flow (w/wo SF) in Chapters 15.1.1.3/15.1.2.3.

55

Analyses Bounded by Another Analysis

  • Chapters 15.2.1.1/15.2.2.1: Loss of External Load (w/wo SF)

- Bounded by the Loss of Condenser Vacuum (LOCV) (w/wo SF) in Chapters 15.2.1.3/15.2.2.3.

  • Chapters 15 15.2.1.2/15.2.2.2:

2 1 2/15 2 2 2: Turbine Trip (w/wo SF)

- Bounded by the LOCV (w/wo SF) in Chapters 15.2.1.3/15.2.2.3.

56

Analyses Bounded by Another Analysis

  • Chapters 15.2.1.4/15.2.2.4: Loss of Normal AC Power (w/wo SF)

- Bounded by the LOCV (w/wo SF) in Chapters 15.2.1.3/15.2.2.3.

- Bounded by the Total Loss of Forced Reactor Coolant Flow in Chapter 15.3.2.1.

57

Analyses Bounded by Another Analysis

  • Chapters 15.3.1.1/15.3.2.2: Partial Loss of Forced Reactor Coolant Flow (w/wo SF)

- Bounded by the Total Loss of Forced Reactor Coolant Flow in Chapter 15.3.2.1.

58

Analyses Bounded by Another Analysis

  • Chapter 15.4.1.5: Chemical and Volume Control System (CVCS)

Malfunction (inadvertent boron dilution)

- Operational Modes 1 and 2

  • Bounded by the CEA Withdrawal at Power in Chapter 15.4.1.3.
  • Bounded by the CEA Withdrawal at Low Power in Chapter 15 15.4.1.2.

412

- Operational Modes 3, 4, 5 and 6

  • Rods are full inserted for Modes 3, 4, 5, and 6.
  • Not N t Impacted I t d 59

A l Analyses Not N t Impacted I t d 60

Analyses Not Impacted

  • Chapter 15.1.3.2: Steam System Piping Failures Inside and Outside Containment (Modes 3 and 4 with All CEAs Fully Inserted)
  • Chapter 15.4.1.4: CEA Misoperation: Single and Subgroup CEA Drop

- No reactor trip generated 61

Analyses Not Impacted

  • Chapter 15.4.1.6: Startup of an Inactive RCS Pump

- Analyzed in Operational Modes 3, 4, and 5.

- CEA SCRAM rods are fully inserted.

  • Ch Chapter t 1515.4.3.1:

4 3 1 Inadvertent I d t t LoadingL di off FFuell A Assembly bl iinto t th the Improper Position

- CEA SCRAM rods do not impact this event.

62

Analyses Not Impacted

- High Pressure Safety Injection (HPSI) system head is less than normal Reactor Coolant System (RCS) pressure

  • Chapter 15.6.3.1: Primary Sample or Instrument Line Break

- Assumes Operator Action for Reactor Trip 63

Analyses Not Impacted

  • Chapter 15.7.3.3: Postulated Radioactive Release Due to Liquid Containing Tank Failures

- Not relevant to this event

  • Chapter Ch 15 15.7.3.4:

7 3 4 Design D i B Basis i FFuell H Handling dli A Accidents id

- Not relevant to this event

  • Chapter 15.7.3.5: Spent Fuel Cask Accidents

- Not relevant to this event 64

Analysis Summary UFSAR Chapter 15

  • Minimal impact on the LOCA safety analyses.
  • Minimal impact on the Non-LOCA safety analyses.

analyses

  • No changes to COLSS ROPM expected.
  • g to CPCS inputs No changes p expected.

p

  • Potential decrease in bounding CEA Ejection physics data.
  • No changes expected to the results and conclusions contained in the current UFSAR Chapter 15.

65

COLSS and CPCS Summary Database and Addressable Constants

  • No changes are expected to the COLSS and CPCS database database.
  • No changes g are expected p to the COLSS and CPCS addressable constants.

66

Summary Overall

  • The Chapter 15 Safety Analyses continue to meet all acceptance criteria
  • The current Licensing Basis will remain bounding for the revised analyses
  • Th The are no changes h tto th the COLSS and d CPCS d databases t b andd addressable constants.

67

License Amendment Request Schedule

  • CEA Drop Time testing is performed at the end of the refueling outage prior to criticality and is typically a critical path activity
  • Failure of the surveillance test would result in an immediate delay in startup following the next refueling outage
  • Westinghouse analysis work is on-going
  • License Amendment Request is expected to be submitted in June 2015
  • Waterford 3 requests the LAR approval be completed by October 25, 2015 68

Conclusion

  • Thank you for your time and consideration
  • Questions?

69

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