L-2011-418, ANP-2903Q1(NP), Revision 0, St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding
| ML11305A088 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 09/30/2011 |
| From: | AREVA, AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| L-2011-418 ANP-2903Q1(NP), Rev 0 | |
| Download: ML11305A088 (16) | |
Text
St. Lucie Unit 1 Docket No. 50-335 L-2011-418 ATTACHMENT 3 Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request NON-PROPRIETARY VERSION (Cover page plus 15 pages)
AREVA NP Inc.
ANP-2903Q1(NP)
Revision 0 St Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding September 2011 A
AR EVA AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903Q1(NP)
Rev. 0 Page i Copyright © 2011 AREVA NP Inc.
All Rights Reserved AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Claddinci ANP-2903Q1(NP)
Rev. 0 ii Nature of Changes Item Page Description and Justification
- 1.
All Responses to NRC RAIs SRXB-41, 48 - 51 AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding Contents 1.0 In tro d u ctio n....................................................................................................................
1 -1 2.0 NRC RAIs with AREVA Responses...............................................................................
2-2 3.0 R e fe re n c e s.....................................................................................................................
3 -1 4.0 D a ta T a b le s....................................................................................................................
4 -1 Tables Table 2-1 Corrosion and Oxidation...........................................................................................
2-2 Table 4-1 Loss of Offsite Power (LOOP) Case Set..................................................................
4-1 Table 4-2 Offsite Power Available (No-LOOP) Case Set..........................................................
4-3 This document contains a total of 15 pages.
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Claddina ANP-2903Q1(NP)
Rev. 0 iv ASI BWR CCTF CE CFR CPHS CSAU DC DEGB DFSS DNB ECCS EFPH EM EPU FA FLECHT-SEASET FP&L Fa Fr HPSI HFP LANL LAR LHGR LOCA LOFT LOOP LPSI MSIV NRC NSSS Nomenclature Axial Shape Index Boiling Water Reactor Cylindrical Core Test Facility Combustion Engineering Inc.
Code of Federal Regulations Containment Pressure High Signal Code Scaling, Applicability, and Uncertainty Downcomer Double-Ended Guillotine Break Design For Six Sigma Departure from Nucleate Boiling Emergency Core Cooling System Effective Full Power Hours Evaluation Model Extended Power Uprate Fuel Assembly Full-Length Emergency Core Heat Transfer Separate Effects and Systems Effects Tests Florida Power & Light Total Peaking Factor Nuclear Enthalpy Rise Factor High Pressure Safety Injection Hot Full Power Los Alamos National Laboratory License Amendment Request Linear Heat Generation Rate Loss of Coolant Accident Loss of Fluid Test Loss of Offsite Power Low Pressure Safety Injection Main Steam Isolation Valve U. S. Nuclear Regulatory Commission Nuclear Steam Supply System AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding v
Nomenclature (Continued)
PCT Peak Clad Temperature PIRT Phenomena Identification and Ranking Table PLHGR Planar Linear Heat Generation Rate PPLS Pressurizer Pressure Low Signal PWR Pressurized Water Reactor RAI Request for Additional Information RAS _.
Recirculation Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RLBLOCA Realistic Large Break Loss of Coolant Accident RV Reactor Vessel RWST Refueling Water Storage Tank SGLS Steam Generator Low (pressure) Signal SIAS Safety Injection Activation Signal SIRWT Safety Injection and Refueling Water Tank SIT Safety Injection Tank SER Safety Evaluation Report THTF Thermal Hydraulic Test Facility w/o Weight Percent AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding 11 1.0 Introduction AREVA performed an RLBLOCA analysis for the EPU cycle of the St. Lucie Unit 1 nuclear power plant. The analysis was performed to support AREVA 14x14 HTP fuel. The results of the RLBLOCA analysis were summarized in Reference 1 and transmitted to FPL for submittal to the NRC for review. What follows in Section 2.0 are AREVA's responses to the NRC's "DRAFT" Request for Additional Information (RAI), questions SRXB-41, plus SRXB-48 through 51.
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding Page 2-2 2.0 NRC RAIs with AREVA Responses RLBLOCA (L-2011-206 Attachment 2 RLBLOCA SR Prop ANP-2903(P) Rev. 1.
SRXB-41:
[2.8.5.6.3.i] Demonstrate that Maximum Local Oxidation and Maximum Total Core-Wide Oxidation will remain below the respective limits of 17.0% and 1.00%
for the entire cycle.
AREVA Response:
The NRC approval of the EMF-2103 realistic LOCA evaluation model (Reference 2) called for the local metal water or oxidation to be reported for the case with the maximum peak cladding temperature. The reasoning employed was that the initialization of the accident simulation with no oxide layer produced an overprediction of the transient oxidation that more than covered the initial corrosion. However, to cover the requirements of GL98-29, the amount of corrosion which bounds the burnup range for the fresh, once-and twice-burnt fuel should be added to the transient oxidation as the metric to compare to the 17 percent local oxidation criteria. To do this the growth of the corrosion layer during the cycle must be accommodated in the calculation.
The values reported in ANP-2903(P) Rev. 1 (Reference 1) were for the limiting case burnup.
Thus, to follow GL98-29 (Reference 3) the following table is provided as supplemental information.
Table 2-1 Corrosion and Oxidation Parameter Fresh Fuel Once burned fuel Twice burned fuel*
Corrosion at the end cycle 1.24%
3.0 %
9.95 %
Transient Oxidation 1.06 %
0.93 %
0.0 %
Maximum Oxidation 2.30%
3.93 %
9.95 %
- Twice burned fuel is sufficiently reduced in energy potential that LOCA significant cladding temperatures and oxidations can not be achieved.
The total core-wide oxidation requirement addresses the release of hydrogen during the LOCA and is, therefore, a limit on the maximum core-wide transient oxidation. The values reported in ANP-2903(P) Rev. 1 are the maximum values of core-wide transient oxidation computed for the case set.
The 0.0209% (Table 3-5, Reference 1) bounds the maximum total core-wide oxidation that would be achieved during a LOCA at any time during the cycle.
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding Page 2-3 SRXB-48:
Please provide a basis for the range of initial conditions for the following: loop flow, pressurizer level, and containment temperature.
AREVA Response:
The RLBLOCA EM (Reference 2) provides a means to broaden the applicability of a single analysis to the support of a plant licensing basis. This is achieved by careful identification of parameter ranges.
Parameter ranges should cover both the normal operational variation allowed and expected measurement and/or other uncertainties. The inputs which were judged, by inference from the Phenomena Identification and Ranking Table (PIRT) or another process, to have a relatively high importance to the calculated results and the use of those results, are sampled in the RLBLOCA analysis. The values for these parameters are modeled as randomly varying quantities, sampled within defined probability distributions. Individual case values for ranged parameters are obtained by sampling the defined probability distribution for each parameter.
The range of the probability distribution includes both the expected or desired operational variation as well as the measurement uncertainty of the parameter value as applicable. The treatment of plant parameters is not rigorously defined in this methodology; rather, any representative distribution is considered acceptable. The choice of how to treat plant-parameter ranges needs to be consistent with the applicability of the analysis in supporting the plants design and control specifications. While the methodology provides for flexibility in describing uncertainty, the analyst is encouraged to apply uniform uncertainty distributions for ranged plant-parameters for the initial application of RLBLOCA to a particular plant. Regulation governing best-estimate LOCA recognizes uniform distributions as statistically conservative.
The loop flow is ranged over the Technical Specification (Table 3.2-1)/COLR (Core Operating Limits Report) minimum departure from nucleate boiling (DNB) flow, 375,000 gpm to the upper limit of 438,500 gpm to bound the maximum RCS flow. The pressurizer level range is 62.6% to 68.6%; this covered the measurement uncertainty of +/- 3% from the nominal pressurizer level of 65.6% for a Tare above 572 OF.
St. Lucie Unit 1 does not have a Technical Specification requirement for pressurizer level.
The containment temperature was ranged in the analysis from 80.5 OF -
124.5 °F, this range is from the Technical Specification maximum (Section 3.6.1.5, including 4.5 OF uncertainty) down to a lower bounding containment temperature value.
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding Page 2-4 SRXB-49:
Please provide the basis for the analyzed single failure assumption. If the basis is generic, provide a St. Lucie 1 specific justification for the use of the generic assumption.
AREVA Response:
The single failure assumption was based off of the approved RLBLOCA EM, EMF-2103(P)(A)
Rev. 0 (Reference 2).
Section 6.5 in ANP-2903 Rev. 1 (Reference 1) provides a detailed discussion of the single failure assumption and a sensitivity study on the limiting case for a maximum ECCS injection scenario.
Due to the early timing of PCT, the maximum ECCS scenario resulted in the same PCT temperature as the RLBLOCA EM single failure, but the quench time and oxidation calculated were greater for the RLBLOCA EM single failure case.
SRXB-50:
Tabulate the initial conditions, operating parameters, PCT, time of PCT, SIT empty time, and safety injection initiation time for all cases (with respect to PCT) both with offsite power and without offsite power. A data file is acceptable (and preferred).
AREVA Response:
The table of requested parameters is found in Section 4.0 at the end of the document.
SRXB-51:
Section 6.1 describes a fuel centerline temperature (FCT) bias. Please address any impacts this has on the cladding surface temperature.
AREVA Response:
ANP-2903(P) Rev. 1 (Reference 1) Section 6.1, Item 1.c, discusses the radial temperature profile of the fuel pellet corrected and uncorrected by the burnup dependant bias and uncertainty (thermal conductivity degradation adjustment). This response states, "As the pellet power is not adjusted the radial temperature profile (for the uncorrected centerline temperature) must follow the corrected profile closely and the two must converge at the surface of the pellet."
This statement is based on the power, the same in both pellets, requiring the same differential temperatures to remove the energy from the pellet surface through the gap, through the cladding, and finally to the coolant.
Because the heat transfer properties in none of these regions are altered by changes in conductivity within the pellet, there is no impact of this bias on the required differential temperatures and the cladding surface temperature.
At accident initiation, the cladding temperature is unchanged by the adjustment in pellet thermal conductivity.
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Claddinq ANP-2903Q1(NP)
Rev. 0 Page 3-1 3.0 References 1
AREVA NP Doc. ANP-2903(P)-001, "St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding," May 2011.
2 AREVA NP Doc. EMF-2103(P)(A), Revision 0, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, April 2003.
3 NRC Information Notice GL 98-29, "PREDICTED INCREASE IN FUEL ROD CLADDING OXIDATION," August 3,1998.
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Claddina ANP-2903Q1 (NP)
Rev. 0 Page 4-1 4.0 Data Tables Table 4-1 Loss of Offsite Power (LOOP) Case Set AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding ANP-2903Q1(NP)
Rev. 0 Page 4-2 Table 4-1 continued
=
AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle ANP-2903Q1(NP)
Realistic Large Break LOCA Summary Report Rev. 0 with Zr-4 Fuel Cladding Page 4-3 Table 4-2 Offsite Power Available (No-LOOP) Case Set AREVA NP Inc.
St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report ANP-2903Q I(NP)
Rev. 0 Page 4-4 Table 4-2 continued AREVA NP Inc.