ML15113A787
ML15113A787 | |
Person / Time | |
---|---|
Site: | Waterford |
Issue date: | 04/22/2015 |
From: | Entergy Operations |
To: | Michael Orenak Office of Nuclear Reactor Regulation |
Wang A | |
References | |
Download: ML15113A787 (69) | |
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- Insertion time is measured from fully withdrawn position to 90% inserted 4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
- Steam Generator replacement
- Reactor Vessel Head replacement
- CEA replacement
- Transition to Next Generation Fuel Product 6
Proposed TS Change
- Waterford 3 TS 3.1.3.4 would be revised to:
- 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
- Updated Final Safety Analysis Report (UFSAR) Chapter 15 Design Basis Events are divided into Loss of Coolant Accident (LOCA) and Non-LOCA transient analyses.
- 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.
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
- Supports COLSS and CPCS
- 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
- 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.
- 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)
- Limiting peak reactor coolant system (RCS) pressure for Moderate Frequency and Infrequent events.
- Limiting peak steam generator (SG) pressure for all DBEs.
- 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
- Supports HFP COLSS ROPM.
- 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.
- 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
- Fuel Centerline Temperature << Fuel Melt Limit
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
- Fuel Centerline Temperature << Fuel Melt Limit
- 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
- Supports COLSS and CPCS.
- 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
- 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.
- 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
- 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
- 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.
- 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
- Revised Reactivity SAI
- Analysis Margin in COLSS HFP ROPM value
- 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.
- 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
margin.
- Revised Reactivity SAI
- COLSS HFP ROPM > Event ROPM
- 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)
- 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
- Large Break LOCA
- SCRAM rod insertion not credited.
- Not Impacted
- Small Break LOCA
- 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
- Diversified SCRAM System setpoints not impacted
- 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)
- 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
- Chapter 15.5.1.2: Inadvertent Operation of the Emergency Core Cooling System (ECCS) during Power Operation
- 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