ML16112A293
| ML16112A293 | |
| Person / Time | |
|---|---|
| Site: | Harris, Robinson  |
| Issue date: | 05/02/2016 |
| From: | Duke Energy Progress |
| To: | Dennis Galvin Plant Licensing Branch II |
| Galvin, D J DORL/LPL2-2 301-415-6256 | |
| References | |
| CAC MF7443, CAC MF7444 | |
| Download: ML16112A293 (30) | |
Text
H.B. Robinson / Shearon Harris Transient Analysis Methodology May 2, 2016 NRC Offices Duke Energy - PWR Methods
Presentation Outline
- Update on Proposed Submittals
- Licensing Approach
- Overview of DPC-NE-3009-P
- Scope of Demonstration Analyses
- Core Physics and Power Distribution Analysis Methodology
- Methodology Changes
- Conclusion Duke / NRC Meeting 2 Duke Energy - PWR Methods
Methods Reports MNS/CNS ONS Proposed RNP/HNP Submittal Date Physics Codes / Models DPC-NE-1005 CASMO-4/SIMULATE-3 DPC-NE-1006 CASMO-4/SIMULATE-3 DPC-NE-1008 CASMO-5/SIMULATE-3 August 19, 2015 Physics Applications Power Distribution Monitoring DPC-NE-2011 NFS-1001 DPC-NE-1002 DPC-NE-2011 revision February 3, 2016 Withdrawn April 7, 2016 Resubmit April/May 2016 Physics Applications Reload Design DPC-NF-2010 NFS-1001 DPC-NE-1002 DPC-NF-2010 revision February 3, 2016 Withdrawn April 7, 2016 Resubmit April/May 2016 NSSS Codes / Models DPC-NE-3000 RETRAN-02 DPC-NE-3000 RETRAN-3D DPC-NE-3008 RETRAN-3D November 19, 2015 Subchannel T/H Methods DPC-NE-3000 DPC-NE-2004 VIPRE-01 DPC-NE-3000 DPC-NE-2003 VIPRE-01 DPC-NE-3008 DPC-NE-2005 (Appendix)
VIPRE-01 November 19, 2015 SCD Methodology DPC-NE-2005 DPC-NE-2005 DPC-NE-2005 revision March 5, 2015 Approved March 8, 2016 Transient Analysis DPC-NE-3001 DPC-NE-3002 SIMULATE-3K (REA)
DPC-NE-3005 SIMULATE-3K (REA)
DPC-NE-3009 SIMULATE-3K (REA)
June/July 2016 Fuel Performance DPC-NE-2008 (TACO-3)
DPC-NE-2009 (PAD 4.0)
DPC-NE-2008 (TACO-3 and GDTACO)
N/A - TS changes only COPERNIC-2 / PAD-5 December 2016 Duke / NRC Meeting 3 Duke Energy - PWR Methods
Licensing Approach
- DPC-NE-3009-P describes the modeling approach used in performing the (U)FSAR Chapter 15 non-LOCA transient analysis
- Extends the Duke DPC-NE-3001-PA / DPC-NE-3002-PA methodology to the Harris and Robinson Nuclear Plants
- System response uses the RETRAN-3D computer code and models described in DPC-NE-3008-P (submitted to the NRC 11/2015)
- DNBR analysis will use the VIPRE-01 models described in DPC-NE-2005-P (approved by the NRC 3/2016) or the extended VIPRE-01 models described in DPC-NE-3008-P Duke / NRC Meeting 4 Duke Energy - PWR Methods
Licensing Approach (cont.)
- LAR submittal
- Intend to submit DPC-NE-3009 as a supplement to the DPC-NE-3008 LAR to avoid linked submittal concerns
- References to DPC-NF-2010 will be provided as an example method that can be used (or an equivalent NRC-approved method)
- Tech Spec 5.6.5 and 6.9.1.6 changes
- (U)FSAR changes
- Implemented via 10 CFR 50.59 following methodology report approval with first in-house reload analysis Duke / NRC Meeting 5 Duke Energy - PWR Methods
Schedule
- Support the reload licensing analysis for Harris Cycle 22 and Robinson Cycle 32
- H1EOC21 (4/18)
- R2EOC31 (9/18)
- Reload Analyses Start:
- HNP (December 2016)
- RNP (Spring 2017)
- Review requested by the middle of 2017 Duke / NRC Meeting 6 Duke Energy - PWR Methods
Overview of DPC-NE-3009-P
- Defines the method used to analyze non-LOCA (U)FSAR Chapter 15 accidents and transients
- Establishes the analysis and modeling assumptions to bound the licensed operating conditions for the current plant design and fuel cycle
- Establishes a set of key physics parameters important to each accident that are verified on a cycle-specific basis
- Methodology used to demonstrate that accident acceptance criteria are satisfied
- Fuel design limits
- System overpressure design limits
- Provides input to dose analysis used to confirm acceptability of dose consequences Duke / NRC Meeting 7 Duke Energy - PWR Methods
Overview of DPC-NE-3009-P (cont.)
- 1.0 Introduction
- 2.0 Background
- 3.0 Simulation Codes and Models
- 4.0 Safety Analysis Physics Parameters Duke / NRC Meeting 8 Duke Energy - PWR Methods
Overview of DPC-NE-3009-P (cont.)
- 5.0 Transient Analysis Methods
- 5.1 Increase in Heat Removal by the Secondary System
- 5.2 Decrease in Heat Removal by the Secondary System
- 5.3 Decrease in Reactor Coolant System Flow Rate
- 5.4 Reactivity and Power Distribution Anomalies
- 5.5 Increase in Reactor Coolant System Inventory
- 5.6 Decrease in Reactor Coolant System Inventory
- 6.0 Demonstration Analyses
- 7.0 Summary Duke / NRC Meeting 9 Duke Energy - PWR Methods
Scope of Demonstration Analyses
- Demonstration analyses completed to provide sample results for various events
- Input assumptions derived from various sources
- Current design inputs, preliminary nuclear design models, proposed input changes, etc.
- Demonstration analyses not intended for direct incorporation into the (U)FSAR
- Demonstration analyses not being submitted for review and approval as new AORs Duke / NRC Meeting 10 Duke Energy - PWR Methods
Scope of Demonstration Analyses (cont.)
Duke / NRC Meeting 11 Scope of Demonstration Analyses Plant/Event Evaluated Harris Robinson 15.1 Increase in Heat Removal by the Secondary System Steam System Piping Failure (MSLB)
Steam System Piping Failure (MSLB) 15.2 Decrease in Heat Removal by the Secondary System Turbine Trip (Submitted for review in separate LAR) 15.3 Decrease in Reactor Coolant System Flow Rate Loss of Flow Locked Rotor 15.4 Reactivity and Power Distribution Anomalies RCCA Ejection Uncontrolled Bank Withdrawal at Power Withdrawal of a Single Full-Length RCCA RCCA Ejection Duke Energy - PWR Methods
Core Physics and Power Distribution Analysis Methodology
- Example analysis methodology is based on the following report:
- DPC-NF-2010, Nuclear Physics Methodology for Reload Design
- DPC-NF-2010 describes Dukes overall reload design methodology and the methodology used to calculate core physics parameters
- An NRC-approved methodology is used to calculate core physics parameters Duke Energy - PWR Methods Duke / NRC Meeting 12
Physics Codes
- CASMO-4/CASMO-5: Develops few-group nuclear data for each unique fuel assembly region in the core
- Cross sections
- Kinetics data
- Assembly discontinuity factors
- Fission product data
- SIMULATE-3: Three-dimensional steady-state core simulator used to calculate reactivity and core power distributions and perform fuel depletion
- SIMULATE-3K: Three-dimensional transient core simulator used for analysis of the RCCA ejection accident
- Model has all of the capabilities of SIMULATE-3 but explicitly accounts for the time-dependent behavior of the neutronics and thermal-hydraulics of each fuel assembly in the reactor core Duke Energy - PWR Methods Duke / NRC Meeting 13
Key Physics Parameters
- Calculation methodology follows approach developed in DPC-NE-3001-PA
- Methodology defines parameters that are important to the transient response for each accident
- Selection objectives
- Bound expected reload values at nominal and transient conditions
- Retain margin to account for uncertainty and future core design and/or operational changes
- Minimize impact on reload core design
- Minimize the potential for re-analysis Duke Energy - PWR Methods Duke / NRC Meeting 14
Parameter Validation
- Cycle-specific confirmation is performed for each reload core
- Parameters are calculated assuming a conservative set of initial conditions and compared against the accident analysis assumptions
- Reload calculated values consider the impacts of:
- Power level
- Soluble boron concentration
- Control rod position
- Burnup
- Xenon
- If a non-conservative parameter is found during verification of the safety analysis, then the accidents affected must be re-evaluated or the loading pattern revised Duke Energy - PWR Methods Duke / NRC Meeting 15
Thermal Limits Evaluation
- Core power distributions are compared against departure-from-nucleate-boiling ratio (DNBR) and centerline fuel melt (CFM) limits to confirm accident analysis acceptance criteria are satisfied
- For accidents where DNB and CFM are allowed to occur, the number of fuel rods exceeding their limit is confirmed less than the value assumed in the dose analysis
- Appropriate uncertainties and penalties are applied to the calculated peaking factors prior to comparison to DNBR and CFM limits Duke Energy - PWR Methods Duke / NRC Meeting 16
Selected Methodology Changes (Relative to DPC-NE-3001-PA and DPC-NE-3002-A)
Duke / NRC Meeting 17 Duke Energy - PWR Methods
- Main Steam Line Break
- RETRAN-3D nodalization changes from DPC-NE-3008 to model Main Steam Line Break (MSLB)
- CHF correlations
- RCCA Ejection
- Gap closure model
- Initial gap conductance
- Enthalpy calculation
Overview of the MSLB Analysis
- Event is initiated from a break in the main steam line
- Depressurization of the steam generator produces a rapid reactor coolant system cooldown and depressurization
- Reactivity insertion from the cooldown (due to negative MTC) causes a return to power, and if a stuck rod is assumed, results in high peaking factors which could potentially challenge DNBR and CFM limits
- System analysis is performed with RETRAN-3D using a split reactor vessel model
- Physics inputs are from an NRC-approved nuclear physics method (such as DPC-NF-2010)
Duke / NRC Meeting 18 Duke Energy - PWR Methods
Overview of the MSLB Analysis (cont.)
- Reactivity addition in RETRAN-3D is verified conservative by modeling the core conditions at the minimum DNBR state point in SIMULATE-3
- SIMULATE-3 must be subcritical at the RETRAN-3D state point conditions
- Multi-channel VIPRE-01 model is used with the power distribution at the peak heat flux state point to confirm DNBR is above the CHF limit Duke Energy - PWR Methods Duke / NRC Meeting 19
MSLB Split Reactor Vessel Model Duke / NRC Meeting 20 Duke Energy - PWR Methods
MSLB Secondary System Modeling (HZP Cases)
Duke / NRC Meeting 21 Duke Energy - PWR Methods
MSLB Core Physics Model
- Asymmetric inlet temperature and flow conditions are modeled based on the orientation of the affected loop
- The stuck rod is assumed in the affected region of the core to maximize peaking
- The SIMULATE-3 reactivity prediction is used to verify that the RETRAN kinetics model is conservative
- SIMULATE-3K is used for LOOP case to account for voiding
- SIMULATE-3 power distributions are input to VIPRE-01 to verify the DNBR acceptance criterion is met Duke Energy - PWR Methods Duke / NRC Meeting 22
MSLB VIPRE-01 Model Duke Energy - PWR Methods Duke / NRC Meeting 23
MSLB CHF Correlations Duke / NRC Meeting 24 Duke Energy - PWR Methods
RCCA Ejection Analysis
- Accident is initiated by a failure in the control rod drive mechanism housing
- The control rod is rapidly ejected from the reactor coolant system pressure boundary
- The resulting power excursion is a function of the reactivity worth of the ejected rod
- The key reactor protection trips are:
- High Neutron Flux
- High Neutron Flux Positive Rate
- Over-Power and Over-Temperature Delta-T Duke / NRC Meeting 25 Duke Energy - PWR Methods
RCCA Ejection Analysis (cont.)
- SIMULATE-3K is used to model the transient power response and the high flux and flux rate trips
- Bounding physics parameters are assumed
- Determined some low ejected rod worth cases do not trip on high flux or high flux rate
- T trips must be credited
- Causal factors Include:
- Orientation of control bank rods relative to the excore detectors
- Lack of a high flux positive rate trip (RNP only)
- Event looks like a single rod withdrawal accident except with a hole in the reactor pressure vessel
- RETRAN-3D used to analyze the transient using the power response from SIMULATE-3K Duke / NRC Meeting 26 Duke Energy - PWR Methods
Example Robinson HFP REA Neutron Power vs. Time Duke Energy - PWR Methods Duke / NRC Meeting 27
RCCA Ejection Analysis (cont.)
Duke / NRC Meeting 28 Duke Energy - PWR Methods
RCCA Ejection Analysis (cont.)
Duke / NRC Meeting 29 Duke Energy - PWR Methods
Conclusion
- Minor changes include:
- Computer code upgrades (RETRAN-3D)
- RETRAN nodalization consistent with models in DPC-NE-3008
- REA changes to address plant-specific differences (RETRAN-3D modeling of T trips)
- REA modeling enhancements (VIPRE-01 dynamic gap model, SIMULATE-3K enthalpy calculation)
- Added MSLB CHF correlations for conditions outside HTP correlation range (Modified Barnett, EPRI)