ML20066F467
| ML20066F467 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 02/28/2020 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20066F760 | List: |
| References | |
| L0-0220-69065 | |
| Download: ML20066F467 (56) | |
Text
L0-0220-69065
Enclosure:
"ACRS Subcommittee Presentation: NuScale FSAR Topic- Resolution of Chapter 15 Phase 2 Open Items," PM-0220-69062, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.36(}.0500 Fax 541.207.3928 www.nuscalepower.com
- NuScale Nonproprietary ACRS Subcommittee Presentation NuScale FSAR Topic Resolution of Chapter 15 Phase 2 Open Items March 2-4, 2020 PM-0220-69062
- ~! NUSCALE" Powe, lo* oll l'"mo,*Hnu Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Presenters Ben Bristol Supervisor, System Thermal-Hydraulics Meghan McCloskey Thermal-Hydraulic Analyst Matthew Presson Licensing Project Manager Paul lnfanger Licensing Specialist 2
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Agenda
- NuScale Design objectives and long term shutdown implications
- FSAR 15.0.6 Return to power analysis
- Boron transport
- Design basis ECCS cooling
- Design basis DHRS cooling
- Beyond design basis conditions
- Incorporates NRELAPS v1 .4
- Minor module model update
- DHRS actuation logic changes
- ECCS changes
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Completely Passive Design Basis
- Fundamental design characteristics enable passive design objectives
- Low core power and large RCS volume
- Simple decay heat removal systems
- Actual plant capabilities for heat removal and reactivity control are much more reliable than existing fleet
- Fail safe valve positions activate passive heat removal
- NPM can reach cold shutdown using CRAs alone
- Accommodates reactivity insertion from complete Xenon burnout
- NuScale PDC-27 commitment for all future core designs
- No active safety systems - no requirement for safety related power and or safety operator mitigation actions 4
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Traditional Analysis Limitations
- Design basis events are analyzed considering highest worth CRA fai~s to insert.
- 1 of 16 as opposed to 1 of 53 (AP1000)
- Small core (larger leakage) leads to proportionally more excess reactivity and larger CRA worth for exterior assemblies.
- Application of WRSO is uniquely penalizing for the NuScale design
- Origins of GDCs indicate no intention of application of stuck rod margin for the purposes of long term hold down.
- Redundant system intended to be used to compensate for Xenon -
burnout (GDC 26 and 27).
- Not required for the NuScale design due to CRA worth and natural Boron redistribution phenomena during extended ECCS operation 5
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Boron Addition Considerations
- Same phenomena as typical PWR post LOCA Boron accumulation
- Lack of continuous boron source supports sufficient margin to precipitation limits
- Boron accumulation phenomena enhances long term shutdown margin, except late in cycle
- Consequences of late in cycle loss of SOM at low temperatures and power levels due to very slow Xenon burnout are not a safety concern.
- NuScale conclusion: An active or passive safety Boron addition system does not make the design safer.
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Principle Design Criteria 27
- DCA includes an exemption request from GDC-27
- NPM design does align with precedent based compliance with GDC-27 due to lack of second safety reactivity control system
- Principle Design Criteria 27
- Passive reactor GDC-27 equivalent
- Ensures the safety related reactivity control system is designed to achieve and maintain subcritical core
- Ensures fuel integrity for an extended overcooling in com bi nation with a partial failure of reactivity system (stuck rod) 7 PM-0220-69062
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PDC-27 Clarification
- RAl-9498 Q# 15-9S 1
- Revised PDC language consistent with Staff interpretation of acceptable consequences for the return to power condition.
- FSAR was updated committing to SAFDLs acceptance criteria for all DBEs.
- Ensures an accident with fuel failure does not precede a return to power where additional source term would need to be analyzed.
The reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
Following a postulated accident, the control rods shall be capable of holding the reactor core subcritical under cold conditions with all rods fully inserted.
Following a postulated accident, the control rods shall be capable of holding the reactor core subcritical under cold conditions, )Nithout margin for stuck rods, provided the specified acceptable fuel design limits for critical heat flux v,ould not be exceeded by the return to po)Ner.
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Agenda
- Overview of boron transport analyses and evaluation of N-x reactivity balance conditions
- FSAR 15.0.6 Return to power analysis
- Boron transport
- Design basis ECCS cooling
- Design basis DHRS cooling
- Beyond design basis conditions
- Incorporates NRELAP5 v1 .4
- Minor module model update
- DHRS actuation logic changes
- ECCS changes
- ~! NUSCALE "
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Compliance with PDC-27
- Immediate shutdown is sufficient to protect RCPB and SAFDLs with margin for the worst rod stuck out of the core
- Cold shutdown is achieved with all control rods fully inserted
- Loss of Shutdown Margin Consequences Benign
- Evaluated with single highest worth control rod fully withdrawn
- Critical power level does not challenge DHRS or ECCS heat removal or SAFDLs
- Probability of the combination of conditions that results in a loss of shutdown return to power with a single rod stuck out of the core is small 10 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Return to Power Mechanisms
- Moderator overcooling
- ECCS and DHRS designed to removed decay and residual heat
- Under cold conditions DHRS or ECCS can cause a fairly rapid temperature decrease and increased moderation
- Fission product decay
- Xenon decay causes a slow post shutdown reactivity insertion
- Boron redistribution
- Boiling/condensing systems cause boron redistribution
- Boric acid is not readily volatilized to the vapor phase and would be expected to recondense
- Results in increasing concentration in boiling region and decreasing concentration in condensing region
Conclusion:
Boron redistribution during extended ECCS operation increases SOM in the core (neglected in OCRP analysis) 11 PM-0220-69062 w~~,~f~.~f Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Loss of Shutdown Margin
- FSAR 15.0.6 evaluation of return to power conditions
- Updated method described in FSAR 15.0.6
- Statepoint analysis with SIMULATES
- NRELAPS quasi-steady analysis
- Critical power level at overlap
- SIMULATES uncertainties accounted for
- Control rod ejection with additional stuck rod only analyzed for short-term response
- Updated resu~ts presented in FSAR 15.0.6
- Recriticality precluded for DHRS cooling with riser uncovery
- DHRS cooling with covered riser non-limiting
- ECCS cooling limiting 12 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Loss of SDM Evaluation
- Performed for spectrum of initial conditions and cooling modes
- Critical power level determined using the SIMULATES core model with WRSO
- Performed for a spectrum of boundary conditions (pressure, temperature, flow)
- Margin is reported to the appropriate analytical limit also described in the LOCA LTR 13 PM-0220-69062
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Loss of SDM Evaluation
- Limiting Initial Conditions
- Minimized Boron (Hot Full Power, Eq. Xe, EOC Core)
- Maximized Cooling (Max pool level, min pool temp, biased high SG & DHRS heat transfer coefficient, max ECCS capacity)
- Additional Penalties for MCHFR Evaluation
- Reactivity bias applied to SIMULATES to account for methodology uncertainty (Increases critical power level)
- Conservative local peaking factor applied to core heat flux
- Dynamic return to power factor of 2.0 applied 14 PM-0220-69062
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Equilibrium Power Results 250
,.*
- Covered Riser
,.
- ECCS Actuated
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- Uncovered Riser LL ,. - Critical Power
-225
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- 150 0.25 0.50 0.75 1.00 1.25 1.50 0
Decay Heat ( /oRTP) 15 PM-0220-69062
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Results - Return to Power Analysis
- Core temperature must be <200°F for recriticality
- Increased pool temperature decreases the magnitude of the return to power, with 140°F precluding a recriticality
- Earliest recriticality determined to occur approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> post-scram
- MCHFR for most limiting results non-limiting relative to other events
- Other AOO acceptance criteria met
- Other SAFDLs demonstrated with OCRP conditions bounded by existing analyses developed for the DCA 16 * !
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Control Rod Ejection, GDC 28, and PDC-27
- RAI 964 7/q 15-29
- Return to power analysis
- Performed to demonstrate compliance with PDC-27
- Is bounding with respect to long-term holddown with a single control rod ejected
- Control rod ejection analysis
- Performed to demonstrate compliance with GDC 28
- GDC 28 imposes core design limits distinctfrom reactivity control system capabilities addressed by GDC 27
- REA is non-mechanistically postulated for the purpose of evaluating the consequences of a limiting reactivity insertion event as required by GDC 28
- Extension of PDC 27 to REA not warranted by unique design considerations 17 PM-0220-69062
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Control Rod Ejection, GDC 28, and PDC-27
- GDC 27 has historically not been applied to a rod ejection accident
- GOC 27 is not cited in SRP 15.4.8 or RG 1. 77
- Application of GOC 27 not required in other approved rod ejection methodologies
- Extension of POC 27 to NuScale REA is not warranted by unique design considerations
- Control rod ejection is non-mechanistically assumed in NuScale design
- FSAR 3.9.3.1.2: CROM pressure housing is a Class 1 appurtenance per ASME BPVC, Section Ill, NCA-1271
- As with other Class 1 vessels and appurtenances, gross failure is not considered credible
- CROM nozzles are integral parts of reactor pressure vessel closure head forging
- CROM nozzle to Alloy 690 safe-end welds are full penetration butt welds
- Estimated likelihood of rod ejection, failure of a control rod to insert, failure of boron addition system that could result in return to power is - 1E-10 per year 18 PM-0220-69062
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Agenda
- Overview of boron transport analyses and evaluation of N-x reactivity balance conditions
- FSAR 15.0.6 Return to power analysis
- Boron transport
- Design basis ECCS cooling
- Design basis DHRS cooling
- Beyond design basis conditions
- Incorporates NRELAPS v1 .4
- Minor module model update
- DHRS actuation logic changes
- ECCS changes
- Overall changes in Chapter 15 analysis results FSAR Rev. 2 to FSAR Rev. 4 19 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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ECCS Boron Transport - Context Context for ECCS boron transport analysis:
- Boron precipitation analysis performed as part of ECCS long term cooling analysis
- Boron dilution analysis performed to :
- Evaluate potential for lower boron concentration fluid in core or near core inlet
- Confirm appropriate scope of return to power analysis by demonstrating that core region concentration remains above initial concentration
- Response to RAI 8930 Boron transport governed by:
- boiling in the core
- condensation in the containment vessel 20 PM-0220-69062
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- Method summary for dilution analysis:
- LTC PIRT high ranked phenomena affecting boron transport evaluated
- Control volume approach to analyze transport between regions
- NRELAP5 used to provide volume fluid masses, flow rates as input for boron transport calculation
- Volatility, entrainment calculated separately
- Boron transport calculation performed separate from NRELAP5
- Cotransport out of RCS hot region
- Demonstrate that RCS hot region concentration remains above initial concentration
- Key areas of NRC review:
- Treatment of boron volatility
- Mixing
- CR opened to evaluate scope of scenarios considered
- Additional discussion in closed session
- conservatively model transport between regions:
Boron distribution factors applied to minimize boron transport in, maximize boron 21 PM-0220-69062
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ECCS Boron Transport - Results
- Results summarized in RAI 8930 show core boron concentration remains above initial concentration
- No net core boron dilution is expected even with biased transport assumptions
- More realistic analysis of boron transport indicates boron concentration in RCS core region is 2-3 times the initial concentration at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Core boron concentration remains above initial concentration for at least 7 days.
- Realistically, long term, high boron concentration expected in RCS hot region, with low concentration in RCS cold region, containment
- Recovering the riser and establishing Mode 3 conditions will take multiple deliberate operator actions following appropriate procedures
- Procedures are developed on a site-specific basis (COL commitments 13.5-2 and 13.5-7.)
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Comparison of ECCS and DHRS Conditions ECCS DHRS
- Cooling established by
- Riser uncovery sustained by boiling/condensing mode significant convective heat transfer through the riser wall
- Will tend to redistribute boron into the RCS hot region and RCS level significantly higher than during ECCS operation - top of steam out of RCS cold, CNV regions generator relatively limited condensing potential compared to CNV
- RCS level well below top of the riser When riser remains covered, primary side natural circulation maintained and boron distribution should remain close
- Recovering the riser and to initial well-mixed condition establishing Mode 3 conditions will take multiple deliberate
- Minimal drivers to redistribute operator actions fol lowing boron in RCS compared to ECCS cooling appropriate procedures
- Recovering the riser and establishing Mode 3 conditions will take multiple deliberate operator actions following appropriate procedures Additional discussion of extended DHRS cooling in closed session.
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- ATWS is typically postulated due to common cause failure of l&C systems to generate a reactor trip signal
- Events would easily be resolved by removal of power from CRDMs
- Focus of ATWS analysis is generally limited to short term RPV/Secondary pressurization analysis.
-- Short term effects are not challenging due to small core power, large RCS volume, and large RSV capacity in NuScale module design.
- Analysis of long term effects of an ATWS are less meaningful for risk insight due to low combined likelihood of mechanical CCF of 16 CRAs
- N-3 transients will immediately shutdown similar to design basis events where WRSO is considered (Ch. 15)
- PRA supporting T/H calculations evaluate failure of all CRAs to insert for 72hr coping period.
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- LOCA/IORV Events
- Short term response
- BOC/EOG - Break flow cause sufficient depressurization and void feedback to make core subcritical.
- Long term response (riser uncovery)
- EOG - Equilibrium power level achieved due to balance ECCS cooling with reactivity feedback mechanisms from void, temperature, and Xenon.
- nonLOCA Events
- Short term response
- EOG - Large negative MTG cause a quick stabilization of core power and DHRS heat removal.
- Long term response (no riser uncovery)
- BOC - Equilibrium power level achieved due to balance DHRS cooling with reactivity feedback mechanisms from temperature, Xenon, and Boron accumulation. RSV venting and subsequent boron concentrating identified as important factor in overall reactivity balance.
- EOG - Equilibrium power level achieved due to balance DHRS cooling with reactivity feedback mechanisms from temperature, and Xenon.
- T/H results support the conclusion that no operator action is required to mitigate event and prevent core damage for short or long term ATWS mitigation.
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Conclusions
- Inherent design characteristics provide ample safety
- Low core power, large RCS inventory, small high pressure containment, and large ultimate heat sink
- Compliance with intent of GDCs is demonstrated for reactivity control systems
- Conservative analysis of the low probability return to power condition demonstrates safety margin
- Boron redistribution is evaluated and demonstrated to not be a safety concern
- Naturally accumulating boron in the core adds to shutdown margin for design basis event and severe accidents.
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Agenda
- Overview of boron transport analyses and evaluation of N-x reactivity balance conditions
- FSAR 15.0.6 Return to power analysis
- Boron transport
- Design basis ECCS cooling
- Design basis DHRS cooling
- Beyond design basis conditions
- Incorporates NRELAP5 v1 .4
- Minor module model update
- DHRS actuation logic changes
- ECCS changes
- ! NUSCALe *
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Ch 15 Changes FSAR Rev. 2 to Rev. 4
- Results from FSAR Rev. 2 presented to ACRS in June, July 2019 in subcommittee and full committee meetings for Chapter 15
- Changes in FSAR Rev. 3 include
- Update from NRELAPS v1 .3 to v1 .4
- Updated NRELAPS base model input
- More conservative core design input in some cases
- DHRS actuation signal changes, addition of secondary side isolation signal
- ECCS actuation signal changes
- Changes in FSAR Rev. 4 include
- ECCS IAB threshold/release pressure changes 28 PM-0220-69062
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NRELAP5 v1.4
- Modifications made from v1 .3 to v1 .4 were due to routine code maintenance
- 26 specific code Fixes (documented in error reports) with most notable being:
- Condensation correlation error corrections
(< 2 psi increase in CNV pressure calculations)
- Correction to choking model quality factor (little to no impact)
- Updated Windows executable to 64-bit version (not used for production calculations)
- 5 new Features - None of which impact DCA calculations
- Added proprietary classifications marking to source files
- Expanded number of elements allowed in water property file (no water property file update)
- Interpolation update for CHF correlation not used in DCA calculations
- Added warning message to users if mass error stop (1 o/o) is disabled
- Removal of Developmental Options from user access 29 PM-0220-69062
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NRELAP5 Base Model
- Revision O released 12/2015 (DCA submittal 12/2016)
- Revision 1 released 8/2017
- Updates for design consistency
- Minor geometry changes based on drawing updates
- Minor RCS flow loss updates (changes in best estimate values)
- Updates for analysis consistency and ease of downstream use
- Minor nodalization changes to match LOCA model
- Added passive heat structures defined in LOCA model
- Other changes
- Change from elevation based to volume based calculation of collapsed liquid level
- Error correction when specifying lower CNV material (had been previously corrected in impacted analysis calculations)
- Revision 2 released 01/2019 (FSAR Rev. 3 submittal 8/2019)
- Removed ECCS actuation on RCS riser level signal
- Minor RCS flow loss updates
- Minor geometry error corrections 30 PM-0220-69062
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Neutronics Range Changes
- For FSAR Rev 3, analyzed more bounding ranges of core design inp~t, including 2 additional depletions for high and low flow rates.
- Different parameter ranges included:
- Most negative OTC (from -2.25 pcm/°F to -2.5 pcm/°F)
- Delayed neutron fraction (~en)
- Augmentation factors for asymmetric reactivity events
- No changes to MTC range 31 PM-0220-69062
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DHRS Actuation Changes
- Summary of change:
- Add secondary side isolation actuation for range of signals that indicate upset in normal secondary side cooling conditions
- DHRS actuation limited to subset of signals indicating insufficient secondary side cooling
- DHRS actuated following secondary side isolation
- Purpose .of change: Support expected plant startup progressions
- Effect of change on transient analyses:
- Heatup events - No change to expected DHRS actuations on high pressurizer pressure or high RCS hot temperature
- Cooldown events - Secondary side isolation may be actuated first; DHRS actuated afterwards on high steam pressure
- Reactivity events, inventory increase, inventory decrease events not significantly impacted 32 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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DHRS Actuation Changes FSARRev.2 FSAR Rev. 3, Rev. 4 DHRS actuation on: SSI actuation on:
- High pressurizer pressure - High pressurizer pressure
- High RCS hot temperature - High RCS hot temperature
- High CNV pressure - High CNV pressure
- Low pressurizer pressure - Low-low pressurizer pressure
- Low-low pressurizer level - Low-low pressurizer level
- Low main steam pressure - Low main steam pressure
- Low-low main steam pressure - Low-low main steam pressure
- High main steam pressure - High main steam pressure
- High main steam superheat - High main steam superheat
- Low main steam superheat - Low main steam superheat
- High under bioshield temperature - High under bioshield temperature
- Low AC voltage - Low AC voltage -
DHRS actuation on:
- High pressurizer pressure
- High RCS hot temperature
- High main steam pressure
- Low AC voltage 33 PM-0220-69062
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DHRS Actuation Changes
- Example impact on cooldown event:
Decrease in FW Temperature MCHFR Case Event Time (sec) Time (sec)
FSARRev.2 FSAR Rev. 3, 4 \
Feedwatertemperature begins to 0 0 decrease Feedwatertemperature reaches 100°F 160 86 High RCS hot temperature limit 125 184 reached High reactor power limit reached 131 187 Reactor trip (high reactor power) 133 189 DHRS actuation (high RCS hot temp) 133 192 SSI actuation (high RCS hot temp) n/a 192 Limiting case occurs where high power, high RCS hot temperature occurs -
same time 34 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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DHRS Actuation Changes
- Example impact on cooldown event:
Increase in Steam Flow Rate MCHFR Case Event Time(sec) Time(sec)
FSARRev.2 FSAR Rev. 3, 4 Steam flow begins to increase 0 p High RCS hot leg temperature reached 60 n/a High reactor power limit reached n/a 63 Reactor trip 68 65 Low pressurizer pressure limit reached n/a 123 SSI actuation n/a 125 (low pressurizer pressure)
High steam pressure n/a 1692 DHRS actuation 68 1697 Maximum power in both cases - 200 MW Limiting case occurs where high power, high RCS hot temperature occurs - same time 35 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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DHRS Actuation Changes
- Example impact to heatup event:
FWLB Limiting DHRS Case Event Time (sec) Time(sec)
FSARRev.2 FSAR Rev. 3, 4 Large FW line break inside CNV 0 0 High CNV pressure limit reached 1 1 RTS actuated 3 3 (high CNV pressure)
Secondary system isolation actuated n/a 3 (high CNVpressure)
High pressurizer pressure limit reached { does not cause additional 7 actuations}
DHRS actuation 3 9 (high CNV pressure) (high PZR pressure)
RSV lift point reached 25 25 DHRS actuation valves open 33 39 36 PM-0220-69062
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DHRS Actuation Changes
- Example impact on reactivity event:
Uncontrolled bank withdrawal at power - MCHFR case Event Time (sec) Time(sec)
FSARRev.2 FSAR Rev. 3, 4 CRA bank begins to withdraw 0 0 High RCS hot temperature limit 178 144 reached High pressurizer pressure limit reached 184 150 Reactor trip actuated 186 152 SSI actuated n/a 152 DHRS actuated 186 152 Limiting case occurs where high power, high RCS hot temperature occurs -
same time when different signal delays are accounted for 37 PM-0220-69062
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I ECCS Valve Operation
- Emergency core cooling system (ECCS) valves receive actuation demand on:
- ECCS actuation signal on high CNV level, or
- Loss of DC power to ECCS trip valves
- Inadvertent actuation block (IAB) feature prevents ECCS valve opening if the differential pressure between the reactor coolant system (RCS) and containment (CNV) is above the IAB threshold pressure
- This feature prevents opening due to spurious signals or equipment failures at normal operating pressures but perm its opening in loss-of-coolant accident (LOCA) conditions
- If IAB is actuated by ECCS demand at high differential pressure, IAB releases at lower pressure and then ECCS valves open 38 PM-0220-69062
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ECCS Actuation Changes FSAR Rev 2 FSAR Rev 4
- ECCS actuated on
- ECCS actuated on \
- High CNV level
- High CNV level (220-260 in) (264-300 in)
- Low RCS riser level (350-390 in)
- Loss of DC power to valve
- Loss of DC power to valve actuators actuators
- RCS riser level is post-accident
- RCS riser level is post-accident monitoring Type B and Type C monitoring Type B and Type C variable variable
- 4 total divisions of RPV riser level
- 4 total divisions ofRPVriser level
- IAB threshold/release
- IAB threshold/release
- Threshold:
- Threshold:
Block if ECCS actuated above Block if ECCS actuated above threshold pressure that is in the 1300 psid; range of 1000-1200 psid does not block below 900 psid
- Release: If IAB blocks, release
- Release: If IAB blocks, release at between 1000-1200 psid 950 psid +/- 50 psi 39 PM-0220-69062
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ECCS Changes - Revised FSAR Analyses
- Impacted FSAR Sections
- FSAR 6.2 Peak CNV Pressure
- FSAR 15.6.5 Loss of Coolant Accidents
- FSAR 15.6.6 ~nadvertent Operation of ECCS
- Revised assumptions
- Assumes all ECCS valves remain closed due to IAB block function above 1300 psid
- Evaluated ECCS valves opening on IAB release pressure between 900 and 1000 psid
- Revised analysis results submitted in September 2019 and reviewed in NRC October audit in Corvallis
- DCA Revision 4, including revised FSAR analysis results, formally submitted January 2020 40 PM-0220-69062
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ECCS Changes - Updated Analysis Results DCARev3 Updated Event / Acceptance Results DCARev4 Comments Criteria Results Peak CNV Pressure Change in peak pressure (RRV Opening) due to staggered IAB 986 psia 994 psia CNV Design release (2"d RRV at 1000 Pressure - 1050 psia psid, RVVs at 900 psid)
LOCA- Minimum Change due to lower IAB Water Level Above 1.7 ft 1.5 ft minimum release pressure Top of Active Fuel 900 psid Change due to model Inadvertent ECCS revisions not IAB threshold valve opening - 1.41 1.32 change MCHFR limit 1.13 41 PM-0220-69062
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Conclusions - ECCS Valve Changes
- CNV peak pressure results slightly more limiting (8 psi) due to explicit evaluation of ECCS valves opening at different IAB release pressures
- LOCA minimum water level above fuel results slightly more limiting (-0.2 feet difference) due to lower minimum IAB release pressure of 900 psid
- Inadvertent ECCS valve opening MCHFR slightly more limiting due to evaluation of error corrections and more bounding model input, not from IAB change
- All updated event results demonstrated margin to acceptance criteria 42 PM-0220-69062
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Agenda
- Overview of boron transport analyses and evaluation of N-x reactivity balance conditions
- FSAR 15.0.6 Return to power analysis
- Boron transport
- Design basis ECCS cooling
- Design basis DHRS cooling
- Beyond design basis conditions
- Incorporates NRELAP5 v1 .4
- Minor module model update
- DHRS actuation logic changes
- ECCS changes
- Overall changes in Chapter 15 analysis results FSAR Rev. 2 to FSAR Rev. 4 43 PM-0220-69062
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FSAR 15 Limiting Transient Results Limiting Limiting Parameter Event Acceptance Criterion Result Result FSAR Rev. 2 FSAR Rev. 4
< 2315 psia(110°/o Pdesign)
Maximum Several or -2170psia - 2170 psia RCS Pressure
< 2520 psia (120%, Pdesign)
I nadve rte nt
< 2315 psia (110% P design) 1582 psia 1592psia Operation DHRS Maximum SG Pressure < 2520 psi a ( 120°/o P design)
SG tube failure 1806 psia 1871 psia Single rod
> 1.284 1.614 1.375 withdrawal MCHFR Inadvertent
> 1.13 1.41 1.32 Opening RRV LOCA > 1.29 1.796 1.74 Level above LOCA > 0 ft 1.5 ft 1.5 ft top of core 44 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Conclusions
- Revised return to power analysis shows ECCS cooling conditions result in equilibrium power at 1-2% RTP
- ECCS boron transport analysis demonstrates that core boron concentration remains higher than initial concentration
- Changes incorporated into FSAR Revision 3:
- Several minor changes in NRELAPS code, NPM plant base model
- DHRS, ECCS actuation changes
- FSAR Ch 15 analysis results demonstrate margin to acceptance criteria 45 PM-0220-69062
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Acronyms AOO -Anticipated Operational Occurrences MCHFR- Minimum Critical Heat Flux Ratio MTC - Moderator Temperature Coefficient CHF-Critical Heat Flux NPM - NuScale Power Module CNV - Containment Vessel OCRP- Overcooling Return to Power COL - Combined License PDC - Plant Design Criteria PIRT- Phenomena Identification and Ranking Table COLR - Core Operating Limits Report RCPB - Reactor Coolant Pressure Boundary CROM - Control Rod Drive Mechanism RCS - Reactor Coolant System REA- Rod Ejection Accident eves - Chemical and Volume Control System SAFDL - Specified Acceptable Fuel Design Limits DHRS - Decay Heat Removal System SOM - Shutdown Margin OTC - Doppler Temperature Coefficient WRSO - Worst Rod Stuck Out ECCS - Emergency Core Cooling System EOC - End of Cycle GDC- General Design Criteria IAB - Inadvertent Actuation Block LCO - Limiting Condition for Operation LOCA - Lossof Coolant Accident 46 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Portland Office Richland Office 6650 SW Red1MJod Lane, 1933 JadVvin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541. 360. 0500 971.371.1592 Charlotte Office Corvallis Office 2815 Coliseum Centre Drive, 1100 NE Circle Blvd., Suite 200 Suite 230 Corvallis, OR 97330 Charlotte, NC 28217 541. 360. 0500 980. 349. 4804 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MD 20852 301. 770.0472 http://vwvw..nuscalepower.com
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Additional Information 48 r PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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MODE Definitions Table 1.1-1 (page 1 of 1)
MODES INDICATED REACTOR REACTIVITY COOLANT MODE TITLE CONDITION (kerr) TEMPERATURES C'F) 1 Operations ~0.99 All:2!420 2 Hot Shutdown <0.99 Any~420 3 Safe Shutdown 1.i> < 0.99 All <420 4 Transition (IIJ(c) < 0.95 NIA 5 Refueling {ell NIA NIA (a) Any CRA capable of withdrawal, any CVCS or CFDS connection to the module not isolated.
(b) All CRAs incapable of withdrawal, CVGS and CFOS connecfio111s to the mcdule isolated, and all reactor vent valves electrically isolated.
(c) All reactor vessel flange bolts fully tensioned.
(d) One or more reactor vessel flange bolts Jess than fully tensioned.
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SG Modeling
- NRELAPS validation and NPM sensitivity calculations:
- SIET-TF1 secondary side heattransfer and pressure drop SIET-TF2 primary and secondary side heat transfer and pressure drop N PM sensitivity calculations for steam generator modeling
- Axial nodalization, heat transfer variation
- Key conclusions for NPM non-LOCA analysis:
Steam generator heat transfer variation affects steady-state conditions:
- Steam generator initial secondary side inventory, steam temperature
- RCS initial flow and temperature conditions due to influence of secondary side conditions on natural circulation driving head Steam generator heat transfer impact on initial conditions affects which process condition is first reached that actuates reactor trip and/or other engineered safety systems
- For events analyzing a spectrum of change, changes in steam generator secondary initial conditions will tend to shift the magnitude of the limiting change but not otherwise change the type of event progression.
Example: the limiting temperature decrease for the decrease in FW temperature event analysis Steam generator heattransferdoes not directly affect margin to MCHFR
- After reactor trip, power decrease much faster than flow rate decrease, even considering variation in steam generator heat transfer 50 PM-0220-69062 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Analysis Results - FSAR Rev. 2; Rev 4 15.1 Increase in heat remova l by secondary system Sec. Event !1 > Peak RCS Pressure Peak SG Pressure MCHFR (Acceptance criteria) (< 110% Pdesign:2310 psia) (< 110% Pdesign: 2310 psia) (> limit: 1.284)
(< 120% Pdesign: 2520 psia) (< 120% Pdesign: 2520 psia) 15.1 .1 Decrease in feedwater temperature 1959 2005 1432 1541 1.921 1.847 15.1 .2 Increase in feedwater 1936 2002 1424 1491 1.944 1.854 flow 15.1.3 Increase in steam flow 2018 1981 1208 804 1.957 1.881 15.1.4 Inadvertent opening of steam generator relief NA NA NA NA NA NA or safety valve 15.1.5 Steam piping failures 2156 2081 1346 1495 1.861 1.866 15.1 .6 Loss of containment vacuum/containment 1992 1937 1342 1426 2.761 2.66 flooding (1 l (1) NuScale unique event Sig nificant margin to acceptance criteria for all events 51 PM-0220-69062 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Analysis Results - FSAR Rev. 2; Rev 4 15.2 Decrease in heat removal by secondary system Sec. Event 11 > Peak RCS Pressure Peak SG Pressure MCHFR (Acee ptance crite ria) (< 110% Pdesign:2310 psia) (< 110% Pdesign:2310 psia) (> limit: 1.284)
(< 120% Pdesign: 2520 psia) (< 120% Pdesign: 2520 psia) 15.2.1 Loss of external load 2158 2161 1474 1545 2.579 2.441 15.2.2 Turbine trip 2158 2161 1474 1545 2.579 2.441 15.2.3 Loss of condenser vacuum 2158 2161 1474 1545 2.579 2.441 15.2.4 Closure of main steam 2160 2161 1481 1512 2.567 2.670 isolation valve 15.2.6 Loss of non-emergency AC 2162 2160 1361 14 15 2.569 2.539 to station auxiliaries 15.2.7 Loss of normal feedwater 2165 2171 1434 1528 2.569 2.426 flow 15.2.8 Feedwater system pipe 2164 2164 1328 1389 2.607 2.496 breaks 15.2.9 Inadvertent operation of the decay heat removal system 2163 2161 1582 1592 2.489 2.67 (1)
( 1) NuScale unique event Si gn ificant margin to acceptance criteria for au events 52 PM-0220-69062 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Analysis Results - FSAR Rev. 2; Rev 4 15 4 R eacf1v1't:van d Power o*ISt rl*bur10n A noma r1es- f OCUS on SAFDL s Sec. Event <1> MCHFR Fuel centerline LHR (Acceptance criteria) (> limit: 1.284) (< Tmelt) (< 21.22 kW/ft) 15.4.1 Uncontrolled control rod assembly withdrawal from subcritical or low power NA
>10 >10 890 8F 1051.8F NA 15.4.2 Uncontrollled control rod assembly withdrawal at 8.97 9.16 power 1 624 1.499 NA NA kW/ft kW/ft 15.4.3 Control rod misalignment 7 10 2.509 1.437 NA NA 8.39 kW/ft 15.4.3 Control rod withdrawal 7 84 1.624 1.375 NA NA 8.29 kW/ft 15.4.3 Control rod drop 8.42 1 641 1.432 NA NA 6.71 kW/ft 15.4.6 lnadi.ertent decrease in boron concentration in RCS NA NA NA NA NA NA 15.4.7 lnadi.ertent loading and operation of a fuel assembly 7 87 in improper position 1 916 1.437 NA NA 8.39 kW/ft 15.4.8 Spectrum of rod ejection accidents NA 2.477 1.838 2162 F 2345F NA Control rod withdrawal has limiting MCHFR for reactivity events 53 PM-0220-69062
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Analysis Results - FSAR Rev. 2; Rev 4 15.5 Increase in reactor coolant inventory Sec. Event <1 > Peak RCS Pressure Peak SG Pressure MCHFR (Acceptance criteria) (S 110% Pdesign: 2310 psia) (S 110% Pdesign: 2310 psia) {i:!: limit: 1.284) 15.5.1 Chemical and volume control system malfunction 2130 2160 1418 1430 2 379 2.702 Significant margin to acceptance criteria 54 PM-0220-69062 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Analysis Results - FSAR Rev. 2; Rev 4 15.6 Decrease in reactor coolant inventory Sec. Event<1> Peak RCS Pressure Peak SG Pressure MCHFR Additional (Acceptance criteria) (< 110% Pdesign: 2310 psia) (< 110% Pdesign : 2310 psia)
(< 120% Pdesign: 2520 psia) (< 120% Pdesign: 2520 psia) 15.6.1 Inadvertent opening of NA NA NA NA reactor safety valve 15.6.2 Failure of small lines carrying primary coolant outside 2047 2067 1368 1473 NA Note 2 containment 15.6.3 Steam generator tube failure 2073 2158 1806 1871 NA Note 2 15.6.5 Loss of coolant accidents 1.796 1.74 1.5 ft 1.5 ft resulting from a spectrum of postulated piping breaks NA NA Minimum level v.4th in the reactor coolant Acceptance above top of pressure boundary criteria : > 1.29 core 15.6.6 Inadvertent operation of Result: Result:
emergency core cooling 1.41 1.32 system <1l NA NA NA Acceptance criteria: > 1.13 (1) NuScale unique event (2) Mass release and iodine spiking time provided as input to radiolog ical analyses SG tube failure maximum secondary pressure remains below design pressure Valve opening and LOCA events demonstrate margin to acceptance criteria 55 PM-02 20-6 9062
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