ML20054A962
ML20054A962 | |
Person / Time | |
---|---|
Site: | NuScale |
Issue date: | 02/19/2020 |
From: | NuScale |
To: | Office of Nuclear Reactor Regulation |
References | |
LO-0220-68854 PM-0220-68852, Rev 0 | |
Download: ML20054A962 (24) | |
Text
NuScale Nonproprietary ACRS Subcommittee Presentation NuScale Topical Report Non-Loss-of-Coolant Accident February 19, 2020 PM-0220-68852
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Presenters Ben Bristol Supervisor, System Thermal Hydraulics Meghan McCloskey Thermal Hydraulics Analyst Matthew Presson Licensing Project Manager Paul lnfanger Licensing Specialist 2
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Outline
- Scop_e of non-LOCA LTR
- Non-LOCA events
- Events and acceptance criteria
- Interface to other methodologies
- Factors controlling margin to acceptance criteria
- Development of non-LOCA EM
- PIRT and gap analysis
- Focus of NRELAPS validation for non-LOCA
- General event analysis methodology
- Specific event analysis 3
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Scope of Non-LOCA Topical Report In Scope Out of Scope
- NRELAP5 system
- Accident radiological dose
- Interface to subchannel analysis and accident radiological
- Control rod ejection analysis
- LOCA and valve opening
- Short-term transient events progression with DHRS
- Peak containment cooling pressure/tern perature analysis
- Long term transient progression with DHRS Riser uncovery Return to power 4
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Non-LOCAEM EM applicable to NuScale Power Module plant design Applicable initiating events:
- Cooldown events
- Reactivity events
- Decrease in FW temperature Uncontrolled bank withdrawal from subcritical
- Increase in FW flow
- Uncontrolled bank withdrawal at power
- Increase in steam flow Inadvertent opening of SG relief or safety valve - Control rod niisoperation
- Steam piping failures (postulated accident)
- Single rod withdrawal
- Loss of containment vacuum Containment flooding
- Control rod drop
- Inadvertent decrease in RCS boron concentration
- Heatup events
- Loss of external load Turbine trip
- Inventory increase event
- Loss of condenser vacuum - eves malfunction
- Closure of MSIV
- Loss of non-emergency AC power
- Inventory decrease events
- Loss of normal FW flow Feedwater system pipe breaks (postulated accident) - Small line break outside containment (infrequent event)
Inadvertent operation of DHRS.
- Steam generator tube failure (postulated accident)
NuScale unique event 5
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Non-LOCA *Event Acceptance Criteria AOO Infrequent Event Accident Description Analysis Acee ptance Criteria Acee ptance Criteria Acee ptance Criteria Reactor Coolant Non-LOCA System Pressure S 110% of Design S 120% of Design S 120% of Design NRELAPS (Pdesign= 2100 psia)
Steam Generator Non-LOCA Pressure S 110% of Design S 120%-of Design S 120% of Design
- NRELAPS (Pdesign= 2100 psia)
Minimum If limit exceed, If limit exceed,
> Limit Subchannel Critical Heat Flux Ratio fuel assumed failed <1> fuel assumed failed <1>
Maximum Fuel If limit exceed, If limit exceed,
< Limit Subchannel Centerline Temperature fuel assumed failed <1> fuel assumed failed (1>
< Limits < Limits < Limits Containment Containment Integrity (pressure, temperature) (pressure, temperature) (pressure, temperature) P/T analysis Escalation of an AOO to an accident (AOO)
If other or No No No acceptance Consequential loss of criteria are met system functionality (IE or accident)?
Normal or Normal Radiological Dose < Limit < Limit Accident Operations radiological (1) NuScale safety analysis methodologies developed to demonstrate fuel cladding integrity maintained.
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Evaluation Models - General Non-LOCA Approach r---------------, ~--------------~
I I I I
Plant design, I
I I
I I M&E releases I
Core design, I
I NRELAPS I
I I VIPRE-01 from T/H I - I I_ response, other Fuel rod design, I ,- system T/H T .,. subchannel Plant initial conditions, response analysis input SSC performance
i I RCS pressure, Accident secondary Fuel cladding radiological pressure, integrity analysis Safe stabilized condition ,,
Radiological Non-LOCA topical report Subchannel topical report TR-0516-49416-P TR-0915-17564-P-A dose
~--------------~ acceptance criteria Accident source term topical report TR-0915-17565-P 7
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Non-LOCA Events -
Margin to Acceptance Criteria Design characteristics governing non-LOCA event transient response and margin to acceptance criteria
- MCHFR: Limited by combination of high power, high pressure, high temperature conditions occurring around time of reactor trip, for reactivity insertion events
- Primary pressure: Protected by RSV lift
- Secondary side pressure: Limited by primary side temperature conditions
- Radiological release: MPS designed to rapidly detect and isolate based on measured conditions
- Establishing a safe, stable condition: MPS designed to trip, actuate DHRS to protect adequate inventory in at least 1 steam generator 8
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Non-LOCA EM Development
- Non-LOCA evaluation model developed to perform conservative analyses , following intent of the RG 1.203 EMDAP and applying a graded approach
- Element 1 - Establish applicable transients and acceptance criteria, develop non-LOCA PIRT
- Element 2, 3, 4
- Leverage NRELAPS development, NRELAPS assessments performed during LOCA evaluation model development.
- Gap analysis performed to evaluate how high ranked phenomena are addressed
- Focused on differences in high ranked PIRT phenomena between LOCAand non-LOCA
- Additional NRELAP5 code validation performed focused on DHRS and integral non-LOCA response
- Suitably conservative initial and boundary conditions applied for non-LOCA analyses
- Sensitivity calculations used to demonstrate factors controlling margin to acceptance criteria 9
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Non-LOCA PIRT Development Event Types SSCs Considered in PIRT Increased heat removal Reactor coolant system 1\/lain feedwater system Decreased heat removal Containment vessel 1\/lain steam system Reactivity anomaly Decay heat removal system Chemical volume control system Increase in RCS inventory Reactor pool Containment evacuation Steam generator tube failure system Phase Identification RCS Response DHRS Operation
- PIRT Figures of merit 1 pre-trip transient higher flow levels at full inactive CHFR power levels RCS pressure 2 post-trip transitional flow levels at startup CHFR transition transitioned power levels RCS, secondary, containment pressures 3 stable natural lower flow levels at decay fully effective CHFR circulation power levels RCS mixture level Subcriticality
- If DHRS actuated by protection system
- Different non-LOCA events involve different plant systems and responses
- Short-term response divided into 3 generic phases with associated FoM e
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NRELAPS Applicability for Non-LOCA After non-LOCA PIRT developed, Key areas identified from gap analysis for short-term non-LOCA analysis:
gap analysis performed to
- DHRS modeling and heat transfer determine how to address high-
- NRELAPS validation against KAIST tests; ranked phenomena: NIST-1 SETs HP-03, HP-04
- NPM sensitivity calculations
- Validation performed as part of NRELAPS assessment for LOCA
- Steam generator modeling and heat transfer evaluation model
- NRELAPS validation against S IET-TF 1, S IET-TF2 tests
- Additional validation or benchmark for non-LOCA - NPM sensitivity calculations
- Reactivity event response
- Conservative input
- NRELAPS benchmark against RETRAN-3D
- Subchannel analysis
- NPM non-LOCA integral response
- NRELAPS validation against NIST-1 IETs NLT-2a, NLT-2b, NLT-15p2 Overall conclusion: NRELAP5 code, with NPM system model, is applicable for calculation of th e NPM non-LOCAsystem response 11 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Non-LOCA Analysis Process Topical report Section 4
- 1. Develop plant base model 5. Identify cases for subchannel NRELAP5 input (geometry, control analysis and extract boundary and protection systems, etc) conditions (if applicable)
- 2. Adapt NRELAP5 base model as - Conservative bias directions:
necessary for specific event
- Maximum reactor power analysis and desired initial
- Maximum core exit pressure conditions
- Maximum core inlet temperature
- 3. Perform steady state and transient
- Minimum RCS flow rate analysis calcufations with - NRELAP5 CHF calculations for NRELAP5 dummy hot rod may be used as a screening tool to assist analysts in
- 4. Evaluate results of transient determining limiting cases to be analysis calculations: evaluated in downstream subchannel
- Confirm margin to maximum RCS analysis pressure acceptance criterion
- 6. Identify cases for radiological
- Confirm margin to maximum SG pressure analysis (if applicable) acceptance criterion
- Maximum mass release case
- Confirm appropriate transient run ti~~
execution to demonstrate safe, stab1l1zed - Maixmum iodine spiking case condition achieved 12 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Non-LOCA Methodology General Methodology Event-specific Methodology (Section 7 .1 ): (Section 7 .2)
Steady-state conditions
- Description of event initiation and progression Treatment of plant controls
- Acceptance criteria 'of interest' Loss of power
- Limiting single failure, loss of Single failure power scenarios , or need for Bounding reactivity sensitivity calculations parameter input
- Initial condition biases and Biasing of other parameters: conservatisms, or need for initial conditions , valve sensitivity calculations characteristics, analytical
- Tabulated representative results limits and response times of sensitivity calculations Operator action Example analysis results provided in Section 8 13 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Conclusions
- Non-LOCA system transient evaluation model developed following a graded approach in accordance with guidance provided in RG 1.203
- Applies to NPM-type plant design natural circulation water reactor with helical coil SG and integral pressurizer
- NRELAP5 used to simulate the system thermal-hydraulic response
- Demonstrate primary and secondary pressure acceptance criteria are met
- Demonstrate safe, stabilized condition achieved
- System transient results provide boundary conditions to downstream subchannel and radiological analyses 14 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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- AOO-AnticipatedOperational
- MPS - Module Protection System Occurrences
- CNV - Containment Vessel
- MSS - Main Steam System
- CVCS-Chemical and Volume Control System
- NIST NuScale lntegralSystemTest-1
- DHRS - Decay Heat Removal System
- NPM - NuScale Power Module
- PIRT- Phenomena Identification and Ranking Table
- EM - Evaluation Model
- EMDAP- Evaluation Model Development and Assessment Process
- FW- Feedwater
- RSV - Reactor Safety Valve
- RVV - Reactor Vent Valve
- IET - Integral Effects Test
- SET - Separate Effects Test
- KAIST - Korea Advanced Institute of
- SG - Steam Generator Science and Technology
- SSC - Structures, Systems, and
- LOCA - Loss of Coolant Accident Components
- MCHFR- Minimum Critical Heat Flux Ratio 15:- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - -
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Portland Office Richland Office 6650 SW Redoood Lane, 1933 Jaclwn 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://I/INI/W. nuscalepowercom
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Power f or all humankind 16 PM-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.
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Additional Material for Public Presentation Previously presented background material 17 PM-0220-68852
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Power Module Overv:iew Integral Pressurized Water Reactor steam line
- Core, steam generator and pressurizer in one vessel feedwater line
- Integrated reactor design, no large-break loss-of-coolant accidents
- Reactor coolant system operated in single phase (liquid) density driven flow
- Safety decay heat removal systems are passive and fail safe steam generator
- Module protection system designed to automate event mitigation actuations ( no operator actions) 18 PM-0220-68852
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ECCS Emergency Core Cooling System
- ECCS valves open to a boiling/condensing circulation flow path to transfer decay and residual heat to reactor pool
- Liquid from containment vessel enters RCS through reactor recirculation reactor vent valve reactor vent valves valves
- Vapor vented from RCS to containment vessel through reactor vent valves
- Steam condenses on inside surface of containment vessel
- Heat transfer through vessel walls to the reactor pool reactor recirculation reactor recirculation
- Actuation Signals: High CNV valve valve level, 24hr loss of AC power
- Fail safe: ECCS valves open on loss of DC power 19 PM-0220-68852
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Decay Heat Removal System {DHRS)
- Removes heat after FWIVs loss of normal cooling
- Boiling/condensing loop - - DI-R actuation valves
- Two redundant trains re actor pool
- Redundant actuation and isolation valves for each train DI-R pass Ive condenser
- Initiates on :
- Loss of power
- Loss of cooling indication (ESFAS Signal)
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Deterministic Event Mitigation Module Protection Functions Event Mitigation Reactivity Control Increase in heat removal transients
- Reactor trip Secondary Isolation
- CVCS/Demineralized Water Decrease in heat removal transients Isolation
- Reactor trip DHRS Actuation RCS and Secondary Inventory Reactivity and power transients Control
- Reactor trip Demineralized Water Isolation
- Containment Isolation Increase in RCS inventory transients
- Secondary Isolation
- Reactortrip eves Isolation Heat Removal Decrease in RCS inventory transients
- DHRS Actuation
- Reactor trip CNV Isolation
- Reactortrip 21 PM-0220-68852 Revision : 0 Copyright 2020 by NuScale Power, LLC.
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Pressure vs. Tern perature Operat ion Map 1
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- Module protection system (Ch. 7, red)
- Technical specification LCOs (Ch. 16, blue) 22 PM--0220-68852
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Loss of Power - Non-LOCA Event Riser MPS Stable DH RS uncovery Rx trip DHRS cooling establis hed may occur 24 hrs 72 h rs
- Availability of AC power
,..._---,...._ _...___, !,____~__/ I =c available, DC power Non-LOCA
) )) AC , DC power available DHRS event initiation cooling affects MPS Rx trip Stable DH RS cool ing Riser uncovery may 24 hrs ECCS whether ECCS DHRS established valves occur valve 72 A
opening hrs AC power unavailable, ((,__~_____.((_--+-----<()==--
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DC power available Non-LOCA event DHRS cooling ECCS cooling actuate, and initiation what time they Rx trip DHRS Stable DHRS cooling establis hed ECCS valve opening open 24 72 onlAB hrs hrs AC power unavailable, DC power Non-LOCA unavailable DHRS ECCS event cooling cooling initiation 23 PM-0220-68852
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5.1 - Summary Description 620 580 LL 0
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