ML20151A418
| ML20151A418 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/31/1987 |
| From: | Conroy B, Paustian H, Wei P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19302D386 | List: |
| References | |
| NEDO-31462, NUDOCS 8804070009 | |
| Download: ML20151A418 (128) | |
Text
- - - - - - -
""&i'ST AUGUST 1987 OYSTER CREEK NUCLEAR GENERATING STATION SAFER /CORECOOL/GESTR LOCA LOSS OF COOLANT ACCIDENT ANALYSIS H.H.PAUST N
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3 NED0-31462 Class I August 1987 ,
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i OYSTER CREEK NUCLEAR GENERATING STATION SAFER /COREC00L/GESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS P. Wei H. H. Paustian B. F. Conroy Approved: N te d d__
L. D. Noble, Manager Reload Nuclear Engineering-1 Approved: _ N R. Artigas, Mandler Licensing & Consulting Services h
NUCLE AR ENERGY BUSINESS OPERA' IONS
- GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENERAL $ ELECTRIC
NED0-31462 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company respecting informa-tion in this document are contained in the contract between GPUN and General Electric Company for this report, and nothing contained in this' document shall be construed as changing the contract. The use of this information by anyone other than CPUN or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
11 l
NEDO-31461 CONTENTS
?*KL 1.0 IhTRODUCTION 1-1
2.0 DESCRIPTION
OF MODEL 2-1 2.1 LOCA Analysis Computer Codes 2-1 2.1.1 SAFER 2-1 2.1.2 GESTR-LOCA 2-1 -
2.1.3 CORECOOL 2-2 3.0 AhALYSIS PROCEDURE 3-1 3.1 BWR/2 Generic Analysis 3-1 3.2 Oyster Creek Specific Analysis 3-1 3.2.1 Break Spectrum Evaluation 3-1 3.2.2 Fuel Exposure Considerations 3-2 4.0 INPUT TO ANALYSIS 4-1 4.1 Plant-Specific Parameters 4-1 4.2 Timing for the Onset of Boiling Transition 4-1 5.0 PLANT-SPECIFIC RESULTS 5-1 5.1 Bresk Spectrum Calculations 5-1 5.1.1 Recirculation Line Breaks 5-1 5.1.2 Non-Recirculation Line Breaks 5-1 5.2 Technical Specification MAPLHGR Limits 5-2 5.3 Alternate Operating Mode Considerations 5-3 5.3.1 Four-Loop Operation 5-3 5.3.2 Reduced Core Flow Operation (ELLLA) 5-4 6.0 CohCLUSIONS 6-1
7.0 REFERENCES
7-1 APPENDIX A - OYSTER CREEK SYSTEM RESPONSE CURVES A-1 APPEhDIX B - BD321B FUEL BUNDLE DESCRIPTION B-i s
lii/iv 1
NEDO-31468 LIST OF TABLES Table Title Page 3-1 Analysis Assumptions for Oyster Creek Calculations 3-3 4-1 Operational and ECCS Parameters 4-2 4-2 Single-Failure Evaluation for Oyster Creek 4-4 5-1 Summary of Recirculation Line Break Result * - Nominal 5-5 Evaluation 5-2 Summary of Recirculation Line Break Results - Appendix K 5-6 Evaluation 5-3 Summary of Non-Recirculation Line Break Results - 5-7 Nominal Evaluation 5-4 MAPLHGR vs. Average Planar Exposure, Five-Loop Operation
- a. P8tRB2992 Fuel 5-8
- b. P8DRB2991A Fuel 5-9
- c. P8DRB265H Fuel 5-10
- d. F8DRB239 Fuel 5-11
- e. BD321B Fuel 5-12 5-5a Summary of Four-Loop MAPLHGR Hultipliers Evaluation 5-13 5-5b Four-Loop MAPLHGR Multipliers for F8x8R and GE8x8EB Fuel Types 5-14 A-1 Oyster Creek Recirculation Line Break Figure Summary A-2 A-2 Oyster Creek Non-Recirculation Line Break Figure Summary A-4 B-1 Fuel Bundle Information for BD321B B-1 v/vi
NED0-31463 I l
e LIST OF FIGURES Figure Title Page 2-1 Flow Diagram of BWR/2 LOCA Analysis Using SAFER 2-3 3-1 Normalized Power (Appendix K) 3-4 5-1 Nominal and Appendix K LOCA Recirculation Line Break Spectrum Comparison 5-15 A-1 DBA DSCG (Nominal) -
a Water Level in Hot and Average Channel A-5 b Reactor Vessel Pressure A-6 c Peak Cladding Temperature A-7 d Heat Transfer Coefficient A-8 A-2 DBA DSCG (Appendix K) -
a Water Level in Hot and Average Channel A-9 b Reactor Vessel Pressure A-10 c Peak Cladding Temperature A-11 d Heat Transfer Coefficient A-12 A-3 DBA Suction (Nominal) -
a Water Level in Hot and Average Channel A-13 b Reactor Vessel Pressure A-14 c Peak Cladding Temperature A-15 d Heat Transfer Coefficient A-16 A-4 80% DBA DSCG (Nominal) -
a Water Level in Hot and Average Channel A-17 b Reactor Vessel Pressure A-18 c Peak Cladding Temperature A-19 d Heat Transfer Coefficient A-20 A-5 801 DBA DSCG (Appendix K) -
a Water Level in Hot and Average Channel A-21 b Reactor Vessel Pressure A-22 c Peak Cladding Temperature A-23 d Heat Transfer Coefficient A-24 A-6 601 DBA DSCG (Nominal) -
a Water Level in Hot and Average Channel A-25 b Reactor Vessel Pressure A-26 c Feak Cladding Temperature A-27 d Heat Transfer Coefficient A-28 vii
NEDO-31462 i.
l LIST OF FIGURES (Continued) ,.
1 j Fiaure Title ,Page f A-7 60% DBA DSCG (Appendix K) -
- a Water Level in Hot and Avetage Channel A-29 l b Reactor Vessel Pressure A-30 i c Peak Cladding Temperature A-31 i Heat Transfer Coefficient d A-32 A-8 40% LBA DSCG (Nominal) - !
u i j a Water Level in Hot and Average Channel A-33 Reactor Vessel Pressure A-34 b
c Peak Cladding Temperature A-35 d Heat Transfer Coefficient A-36 J
A-9 40% DBA DSCG (Appendix K) -
1 a Water Level in Hot and Average Channel A-37 l l b Reactor Vessel Pressure A-38 ;
i c Peak Cladding Temperature A-39 d Heat Transfer Coefficient A-40 A-10 1.0 Ft2 DSCG (Nominal) - ;
i
- a Water Level in Hot and Average Channel A-41 ;
i b Reactor Vessel Pressure A-42 '
j c Peak Cladding Temperature A-43 l Heat Transfer Coefficient
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i d A-44 i A-11 0.5 Ft2 DSCG (Nominal) -
4 j a Water Level in Hot and Average Channel A-45 b Reactor Vessel Pressure A-46 c Peak Cladding Temperature A-47 l d Heat Transfer Coefficient A-48 '
A-12 0.1 Ft2 DSCG (Nominal) -
i a Water Level in Hot and Average Channel A-49 b Reactor Vessel Pressure A-50 c Peak Cladding Temperature A-51 i d Heat Transfer Coef ficient A-52 l
l j viii i
i
NEDO-31462 LIST OF FICURES (Continued)
Figure Title Page i i
A-13 0.05 Ft2 DSCG (Nominal) - I a Water Level in Hot and Average Channel A-53 l b Reactor Vessel Pressure A-54 e Peak Cladding Temperature A-55 d Heat Transfer Coefficient A-56 A-14 DBA DSCG - High Exposure (Nominal) - l a Water Level in Hot and Average Channel A-57 b Reactor Vessel Pressure A-58 c Peak Cladding Temprature A-59 d Heat Transfer Coefficient A-60 i A-15 DBA DSCG - High Exposure (Appendix K) - I a kater Level in Hot and Average Channel A-61 l b Reactor Vessel Pressure A-62 c Peak Cladding Temperature A-63 d Heat Transfer Coefficient A-64 e Oxide Thickness A-65 A-16 Core Spray Line (Nominal) -
a Water Level in Hot and Average Channel A-66 b Reactor Vessel Pressuro A-67 c Peak Cladding Temperature A-68 d Heat Transfer Coefficiene A-69 A-17 Steam Line Inside Containment (Nominal) -
a Water Level in Hot and Average Channel A-70 b Reactor Vessel Pressure A-71 e Peak Cladding Temperature A-72 d Heat Transfer Coefficient A-73 A-18 Steam Line Outside Containment (Nominal) - l a Water Level in Hot and Average Channel A-74 b Reactor Vessel Pressure A-75 c Peak Cladding Temperature A-76 d Heat Transfer Coefficient A-77 ix l i l l I -. , , _ - _ .-. - - _ _ _ , _ . -
T ,
NED0-31463 LIST OF FIGURES (Continued) l Pigure Title h !
l A-19 Feedwater Line (Nominal) -
a Water Level in Hot and Average Channel A-78 b .eactor Vessel Pressure A-79 c Peak Cladding Temperature A-80 d Heat Transfer Coefficient A-81 B-1 Enrichment Distribution for the BD321B Fuel Bundle B-2 B-2 Gadolinium Distribution for the BD321B Fuel Bundle B-3 l
l X l 4
NEDo-31462 1.0 INTRODUCTI0h The purpose of this document is to provide the results of the loss-of-coolant accident (LOCA) analysis for the Oyster Creek Nuclear Generating Station. The analysis was performed using tLe NRC approved General Electric (GE) SAFER LOCA code and application methodology for BWR/2 plants.
This analysis of postulated plant LOCAs is provided in accordance with NRC requirements and demonstrates conforcance with the ECCS acceptance cri-teria of 10CFR50.46. The objective of the LOCA analysis contained herein is to provide assurance that the most limiting brea': size, break location, and single failure combination has been considered for the plant. The require-ments for demonstrating that these objectives have been satisfied are given in Reference 1. The documentation contained in this report is intended to satisfy these requirements.
A description of the LOCA models and their application is contained in Reference 2. The Oyster Creek values of the peak cladding temperature (PC7) and maximum oxidation fraction for use in liceesing evaluations are calculated for the limiting break. The results conform to all the requirements of 10CFR50.46 and Appendix K.
1-1/1-2
NED0-31462
2.0 DESCRIPTION
OF MODEL The General Electric evaluation model used for the Oyster Creek loss-of-coolant accident (LOCA) analysis consists of three major computer codes.
SAFER performs the long-term water level and inventory calculations and fuel rod heatup calculations with the gap conductance supplied by GESTR-LOCA.
COREC00L is used to analyze the transient af ter the core is uncovered and per-forms detailed evaluations of the core spray and radiation heat transfer and fuel rod heatup in the high power bundle. These models and their application are discussed in Reference 2. Figure 2-1 shows a flow diagram of the usage of these computer codes, indicating the major code functions and the transfer of major data variables.
2.1 LOCA ANALYSIS COMPUTER CODES 2.1.1 SAFER The SAFER code is used to calculato the long-term system response of the reactor for reactor transients over a complete spectrum of hypothetical break sizes and locations. SAFER is compatible with the GESTR-LOCA fuel rod model f or gap conductance and fission gas release. SAFER tracks, as a function of time, the core water level, system pressure response, ECCS performance, and other primary thermal-hydraulic phenomena occurring in the reactor. SAFER realistically models all regimes of heat transfer which occur inside the core during the event, and provides the outputs as a function of time for heat transfer coefficients and FCT. SAFER also provides initial and boundary con-ditions, for the high power fuel bundle, to COREC00L.
2.1.2 CESTR-LCCA The CESTR-LCCA code is used to initialize the fuel stored energy and fuel rod fission gas inventory at the onset of a postulated LOCA for input to S AF ER . GESTR-LOCA also initializes the transient pellet-cladding gap conduc-tance in SAFER, 2-1
NED0-31462 2.1.3 CCREC00L COREC00L is a model for evaluation of core heatup transients for a fuel d
bundle during the period when the core is uncovered. It has detailed core spray heat transfer and thermal radiation models which can provide more realistic predictions of fuel rod heatup at high cladding temperatures (e.g. ,
1700*F). The fuel rod model in COREC00L includes the GESTR transient gap conductance model and the SAFER rod swelling / perforation model.
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NEDO-31462 3.0 ANALYSIS PROCEDURE 3.1 BkR/2 CLNERIC ANALYSIS For the BWR/2 product line, the limiting break was determined from the nominal break spectrum as that break size, location, and ECCS component fail-ure combirstion that yielded the highest nominal PCT. An Appendix K calcula-tion, utilizing the required features of 10CFR50 Appendix K, was performed for the limiting break.
was found to be the limiting break in the nominal break spectrum f or the BWR/2 product line. As a result, this case was used to perform the Appendix K calculation. The results of the Appendix K calculation deconstrate that a discharge coefficient of in the Hoody Slip Flow Model yields the highest calculated PCT.
Comparison of the Appendix K licensing basis and the upper bound (95th percentile) results demonstrated the conservatism of the BWR/? iicensing application methodology.
3.2 OYSTER CREEK SPECIFIC ANALYSIS 3.2.1 Break Spectrum Evaluation The plant-specific analysis performed for Oyrter Creek consisted of break sizes ranging from 0.05 ft 2 to the maximum of a DBA recirculation line break (4.66 ft ). This plant-specific analysis evaluated recirculation line and non-recirculation line breaks, as well as an assesseent of limiting break location and ECCS component failure. The analysis assumptions (nominal and Appendix K) are presented in Table 3-1. The break spectrum evaluation was performed using the F8x8R fuel. The conclusions from this evaluation are also applicable to the GE5x8EB fuel.
3-1
NEDO-31663 First, the various breaks were evaluated using the nominal assumptions.
The case with the highest PCT was determined to be the
, which became the limiting nominal case. The limiting scenario was then analyzed again with specifications for the Appendix X calculation (see table 3-1). The results of the Oyster Creek nominal and Appendix K cases were compared to assure that the PCT trends as a function of break size are I consistent with each other and with those of the generic BWR/2 break spectrum curves (Section 3.1). These results are presented in Section 5.0.
3.2.2 Fuel Exposure Considerations As discussed in Reference 2, the ECCS acceptance criteria of 10CFR50.46 which are most significant to the BkR/2 LOCA analysis require that the cal-culated FCT following a postulated LOCA shall not oxceed 2200*F and that the calculated maximum cladding local oxidation fraction shall not exceed 17%.
For a BWK/2 plant, the LCCS performance is limited by different factors as the fuel exposure increases.
l 3-2
. 1 NED0-31463 Table 3-1 ANALYSIS ASSUMP;10h5 FOR OYSTER CREEK CALCULATIONS Notinal Appendix K
(see Figure 3-1)
- 2. Transition Boiling Iloeje Correlation 300*F Temperature
- 4. Metal-Water Reaction Cathcart Baker-Just
- 5. Core Power 100% 102%
- 6. hAPLHGRa (kW/ft)
Low Exposure High Exposure
- 7. ECCS Water Temperature 120*F 120'F
- 8. ECCS Flow See Table 4-1 See Table 4-1
- 9. ECCS Flow to Hot Bundle (2 Core Sprays)
- 10. Fuel Type P8x8R F8x8R
abased on PLHGR of A multiplier of 1.02 was applied to the Appendix K values.
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NED0-31462 1
4.0 INPUT TO ANALYSIS 4.1 PLANT-SPECIFIC PARAMETERS ,
l A list of the significant Oyster Creek plant-specific input parameters to the LOCA analysis is presented in Table 4-1. Table 4-2 identifies the break locations and-corresponding single-failure / system available combinations :
1 specifically evaluated for Oyster Creek. 1 4.2 TIMING FOR THE ONSET OF BOILING TRANSITION The current Oyster Creek LOCA licensing document (Reference 3) concludes, from the results of the LAMB and SCAT evaluations (based on an initial MCPR of
, that nucleate boiling is maintained prior to core uncovery for small recirculation line breaks . For large breaks (DBA to DBA), where there is very rapid flow coastdown, the duration of nucleate boil-ing following the break is calculated using the GE dryout correlation (in the ,
CHASTE code), which is based on instantaneous flow stagnation conditions. (The I LAMB, SCAT and CHASTE models and applications are described in Reference 4.)
In this Oyster Creek SAFER analysis, for break sizes from DBA to
, the Sreak spectrum evaluation (results summarized in Section 5.0) utilized c timing for the onset of boiling transition based ou the GE dryout correlation. For break sizes smaller than , nucleate boiling was assumed to be maintained until core uncovery.
This approach established that the Oyster Creek SAFER-LOCA licensing evaluation is dependent upon the LAMB and SCAT analyses only for those break sizes less than . Results of the break spectrum evaluation (Section 5.0) indicate that the small break PCTs are significantly below those of the larger recirculation line breaks. Therefore, future poten-tial changes in MCPR limit are not expected to affect the limiting LOCA sce-l nario (and the resultant MAPLHGR calculations), because the evaluation of the large break is independent of the NCPR limit.
4-1
l NELO-31462 l
Table 4-1 OPERATIONAL AND ECCS PARAMETERS A. Plant Parameters Core Thermal Power (MWth)
Nominal 1930 (100% of Rated)
Appendix K 1969 (102% of Rated)
Vessel Steam Output (1bm/hr)
Vessel Steam Dome Pressure (psia)
Maximum Recirculation Line 4.66 Break Area (ft2)
Initial MCPR Initial Water Level B. Emergency Core Cooling System Parameters Core Spray System Assumed System Configuration:
One Loop The Other Loop Vessel Pressure versus System Flow Rates:
One Loop The Other Loop Initiating Signals and Setpoints Low Water Level or High Drywell Pressure (psig) s 4-2
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NED0-31462 Table 4-1 (Continued)
OPERATIONAL AND ECCS PARAMETERS Maximum Allowable Delay Time from Initiating Signal to Pump at Rated Speed (sec)
Injection Valve Stroke Time (sec)
Pressure Permissive at Which Injection Valve Opens (psig)
Core Spray Flow to Hot Bundle (2 headers)
(gpm)
ADS Total Number of Valvec in System Number of Valves Assumed in Analysis humber of Valves Available After Single Failure Minimum Flow Capacity of 3 Valves (1bm/hr) at Vessel Pressure (psig)
Initiating Signals Low Water Level and High Drywell Pressure (psig)
Time Delay After Initiating Signals (sec)
Emergency Condensers Total Number of Emergency Condensers Assumed in Analysis 4-3
. _ . = _ . _. .. .
NED0-31462 Table 4-2 SINGLE-FAILURE EVALUATION FOR OYSTER CREEK Assumed Break Location Single Failure
- Available Systems Recirculation Line Feedwater and Main Steamlines Core Spray Line CS = Core Spray ;
l ADS = Automatic Depressurization System 1
[
- e 4-4
NED0-31462 5.0 PLANT-SPECIFIC RESULTS 5.1 BREAK SPECIRUM CALCULATIONS 5.1.1 Recirculation Line Breaks A sufficient number of break sizes and ECCS failure combinations were evaluated using nominal input conditions. The results (Table 5-1) identified the as limiting. Analyses with Appendix K input assumptions were performed for four break sizes from the limiting sce-nario determined by the nominal break spectrum. Table 5-2 lists the Appendix K PCT results. Figure 5-1 shows a comparison of these two break spectrums and, in both cases, the highest calculated PCT is associated with the largest break area.
is the limiting break for the nominal break spectrum with a calculated peak cladding temperature of . The corresponding PCT for this break with Appendix K specified models was calculated to be and for the , respectively. Plots showing system responses for all break spectrum cases are presented in the Appendix A to this report.
5.1.2 Non-Recirculation Line Breaks 1
l Evaluations were also performed for some of the non-recirculation line breaks. These breaks (including feedwater, core spray and main steam lines) were evaluated with the nominal input conditions and maximum line break sizes.
PCT results (Table 5-3) show that these non-recirculation line guillotine breaks are far from becoming candidates for the limiting event. The same con-clusion applies for break sizes smaller than the guillotine break for these lines. The system responses of these breaks are also presented in the Appendix A.
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NED0-31462 Oyster Creek plant-specific evaluations were not performed for other non-recirculation line breaks (e.g., EC lines, liquid instrument lines, cleanup system lines, etc.). These non-recirculation line breaks will not become can-didates for the limiting event, since they are essentially the same as small recirculation or steam line breaks.
5.2 TECHNICAL SPECIFICATION MAPLHGR LIMITS GE BWR MAPLHGR limits (as a function of fuel exposure) are based on the j l
i For BWR/2 plants, in general, and the Oyster Creek plant, specifically, the MAPLHGR calculated from the is limiting for most of the exposure r'ange and determines the Technical Specification limits.
The MAPLNGR limits for the P8x8R and GE8x8EB fuel bundles were evaluated as a function of exposure with the limiting scenario identified in the break spectrum analyses.
Table 5-4 lists the MAPLNGR limits (five-loop operation) along with the calculated PCT and peak local oxidation fraction for the following bundles:
P8DRB299Z, P8DRB299ZA, P8LRB265H, P8DRB239 and BD321B. The BD321B bundle is described in Appendix B; the other bundles are described in Reference 4.
For fuel types in future Oyster Creek reloads, this SAFER /LOCA report can serve as an evaluation basis for the plant system responses, and supplemental calculations can be performed to determine the fuel type specific MAPLHGRs.
5-2
NED0-31462 5.3 ALTERNATE OPERATING MODE CONSIDERATIONS 5.3.1 Four-Loop Operation l
There are two main differences in the LOCA analysis for four-recirculation-loop operation, as compared to the normal five-loop case:
I The effects of these differences on the SAFER calculations will depend on the break size.
i Evaluation results (considering both P8x8R and GE8x8EB fuel types) for i
four-loop operation, with the inoperative loop isolated, are summarized in l Table 5-Sa. The DBA break and the break (highest small-break PCT from the nominal analysis) cases were calculated with compared with the five-loop 5-3 i
l NEDO-31464.
l base cases. The results show that are adequate to compensate for the effect of loss of inventory and the faster coastdown for the large and small breaks. Table 5-Sb summarizes the four-loop MAPLHGR multipliers for the P8x8 and GE8x8EB fuel.
5.3.2 Reduced Core Flow Operation (ELLLA)
The impact, on MAPLHGR limits, of operating at rated reactor power and reduced core flow [i.e., in the Extended Load Line Limit Analysis (ELLLA) l 4
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NED0-31462 Table 5-1
SUMMARY
OF RECIRCULATION LINE BREAK RESULTS -
NOMINAL EVALUATION (1)
Core-Wide Break 2 PCT ~ Peak Local Metal-Water Size (ft ) (*F) Oxidation (%) Reaction (%)
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NEDo-31462 Table 5-2
SUMMARY
OE RECIRCULATION LINE BRJAK RESULTS -
l APPENDIX K EVALUATIONl11 i
Core-Wide Break 2 PCT Peak Local Metal-Water Size (ft ) ('E) Oxidation-(%) Reaction (%)
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NEDO-31462 Table 5-3
SUMMARY
OF NON-RECIRCUIATION LINE BREAK RESULTS - NOMINAL EVALUATION (
Core Wide PCT Peak Local Metal-Water
('F) Oxidation (%) Reaction (%)
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, NED0-31462 l
Table 5-4a MAPLHGR vs. AVERAGE PLANAR EXPOSURE FIVE-LOOP OPERATION Plant: Oyster Creek Fuel Type: P8DRB299Z Average Planar Exposure MAPLHGR PCT Local (GWd/MTU) (kW/ft) (*F) Oxidation Fraction 0.22 10.2 1.1 10.2 ;
5.5 10.2 11.0 10.2 16.5 10.1 l 20.0 9.0 25.0 8.8 l 27.5 8.8 33.0 8.8 38.5 8.6 l 44.0 8.6 50.0 8.4 5-8 4
- - _ _ ..-.- . _ . . . - . .=.- . . _
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NEDO-31462 Table 5-4b l MAPLHGR vs. AVERAGE PLANAR EXPOSURE l F1VE-LOOP OPERATION 1
Plant: Oyster Creek Fuel Type: P8DRB299ZA Average Planar Exposure MAPLHGR PCT Local (GWd/NTU) (kW/ft) (*F) Oxidation Fractio _n_
0.22 10.2 1.1 10,2 5.5 10.2 11.0 10.2 16.5 10.1 25.0 8.8 27.5 8.8 33.0 8.8 38.5 8.6 44.0 8.6 50.0 8.4 l
5-9
NED0-31462 Table 5-4c l
l MAPLHGR vs. AVERAGE PLANAR EXPOSURE FIVE-LOOP OPERATION
(
l Plant: Oyster Creek Fuel Type: P8DRB265H Average Planar Exposure MAPLMGR PCT Local (GWd/MTU) (kW/ft) (*F) Oxidation Fraction 0.22 10.2 ,
1.1 10.2 5.5 10.2 11.0 10.1 16.5 10.0 25.0 8.8 ;
27.5 8.8 33.0 8.7 38.5 8.7 44.0 8.5 50.0 8.4 l
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NEDO-31462 Table 5-4d MAPLHGR vs. AVERAGE PIANAR EXPOSURE F1VE-LOOP OPERATION Plant: Oyster Creek Fuel Type: P8DRB239
(
Average Planar Exposure MAPLHGR PCT Local (GWd/MTU) ,
(kW/ft) (*F) Oxidation Fraction 0.22 10.2 1.1 10.2 5.5 10.2 11.0 10.2 i
16.5 10.1 25.0 8.8 27.5 8.8 j l
33.0 8.8 38.5 8.7 i 44.0 8.7 50.0 8.1 l
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NED0-31462 Table 5-4e l MAPLHGR vs. . AVERAGE PLANAR EXPOSURE FIVE-LOOP OPERATION Plant: Oyster Creek Fuel Type BD321B l
Average Planar Exposure MAPLEGR* PCT Local (GWd/MTU) (kW/ft) ('F) Oxidation Fraction 0.22 10.4 1.1 10.5 5.5 11.0 11.0 11.5 16.5 11.3 l 25.0 9.6 ,
t 27.5 9.5 ;
33.0 9.5 38.5 9.4 44.0 9.4 l
50.0 8.7 I
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- For exposure , the MAPLHGR limits are based on fuel thermal-mechanical design criteria and not from ECCS considerations.
i 5-12 1 i
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NEDO-31462 l l
l l
Table 5-Sa i
SUMMARY
OF FOUR-LOOP MAPLHGR MULTIPLIERS EVALUATION P8x8R AND GE8x8EB FUEL, ISOLATED CONDITION Loops Parameter Five* Four* l
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Notes l
- Four-loop and five-1 cop are evaluated with identical Appendix K conditions.
l 5-13
l NEDo-31462 l
Table 5-Sb FOUR-LOOP MAPLHGR HULTIPLIERS FOR P8x8R AND GE8x8EB FUEL TYPES 5-14
z
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a l H W O Figure 5-1. Nominal and Appendix K LOCA Recirculation Line Break Spectrum Comparison
NEDD-31462
6.0 CONCLUSION
S The discussion in this section demonstrates compliance with the NRC Safety Evaluation Report (SER), Reference 1, which specifies the necessary ;
conditions for demonstrating application of the approved methodology. These conditions are:
(1) The application methodology is used only with the GESTR/LOCA and SAFER /CORECOOL computer models which have been approved by the NRC.
(2) Both the generic nominal PCT versus break size curve and the generic j Appendix K PCT are shown to be applicable on a plant-specific basis.
(3) For the limiting case using the Appendix K model, the criteria of ,
10CFR50.46 are met, and the Appendix K PCT exceeds the upper bound i j PCT. (The upper bound PCT is evaluated at the 95th percentile of PCTs calculated to occur under the limiting cotdition by adding an )
uncertainty allowance to a best-estimate calculation for that condition.)
The Oyster Creek plant-specific LOCA analysis was performed using the NRC-approved SAFER /COREC00L and GESTR-LOCA codes and the app.ication method-ology described in Reference 2. The recirculation line break results presented in Section 5.1 demonstrate that a sufficient number of Oyster Creek plant-specific PCT points have been evaluated to verify that the trend of the PCI curves, for both the nominal and Appendix K calculations, is similar to :
the (Reference 2) generic PCT versus break size curves. Thus, conditions 1 l
and 2 from above are satisfied.
It has been demonstrated generically that the PCT calculated in accord-ance with the application methodology described in Reference 2 maintains mar-gin for licensing evaluations (i.e., the licensing basis PCT is at least the upper 95th percentile PCT). This was verified by separate calculations to determine the upper 95th probability values of PCT at the most limiting con-ditions. These calculations were performed to qualify the "Appendix K 6-1 I
_ . . _ _ _ _ _ _ . . _ , _ . . __ _ _. ______l
Procedure" as being sufficiently conservative. The generic upper bound PCTs, ;
which include a 50*F conservatism (in AS) assigned by the NRC (Reference 1),
. By compar-ison, the generic licensing basis Appendix K evaluation with SAFER /CCRECOOL ;
for the limiting conditions provided PCTs of margins to the upper bound requirements.
The Oyster Creek plant-specific Appendix K analysis will have similar margin to the 95th percentile PCT because of the following considerations:
(1) Oyster Creek analysis assumes an ECCS configuration similar to that used in the generic BWR/2 analysis, with one exception; the generic l
analysis had operating energency condensers, while the Oyster Creek plant-specific analysis assumed that the emergency condensers are not operable. This assumption has negligible impact on the PCT results of the limiting large break event, due to the very rapid j system blow-down and depressurization rates. Therefore, the limit- [
ing case LOCA for both Oyster Creek and the generic BWR/2 f
i (2) The key operating parameters for the plant-specific Oyster Creek analysis are similar to the inputs used in the calculations of the generic analysis PCT.
(3) The similarity between the generic and the plant-specific evalua-tions (in plant configuration and the operating parameters) is responsible for the sAmilar PCTs calculated with SAFER /COREC00L.
j The generic nominal SAFER /CCREC00L analysis reported (for the limiting scenario), while the plant-specific analysis respectively. The generic Appendix K (licensing basis) SAFER /
COREC00L results were ; the plant-specific results were also very similar at l
l l 6-2 l
. ~ . --
NED0-31462 The similarities in plant configuration and operating parameters and the close correspondence (i.e., less than 20*F variance) of both nominal and Appendix K PCTs between the generic and the plant-specific analyses indicate that the upper bound PCTs would closely correspond also. Therefore, the plant-specific Appendix K PCTs (for the limiting case) are greater than the upper bound PCTs.
Therefore, it is confirmed that the generic assessment (Reference 2) is applicable to Oyster Creek and the Oyster Creek Appendix K licensing basis analysis exceeds the upper bound 95th percentile PCT. Also, the plant-specific results of the Appendix K licensing analysis of Section 5.0 meet the criteria of 10CFR50.46. Thus, condition 3 of the SER requirements is also satisfied.
The f uel exposure-dependent MAPLHGR calculations for individual fuel bundle types are performed based on the limiting case (from the break spectrum evaluation) using Appendix K assumptions. The technical specification MAPLHGR limit is the most limiting of either this ECCS MAPLEGR or the MAPLEGR based on fuel thermal-mechanical design limits.
In conclusion, it is verified that the Oyster Creek plant-specific SAFER /COREC00L/GESTR-LOCA analysis meets the explicit requirements of the l Reference 1 NRC Safety Evaluation Report. l I
l .
l l
6-3/6-4 1
i
- .n . . - - . - , . , ,- -,
7.0 REFERENCES
l
- 1. Letter, A. C. Thadani (NRC) to H. C. Pfefferlen (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-30996-P, Volume II, ' SAFER Model for Evaluation of Loss +of-Coolant Accidents for Jet and Non-Jet Pump Plants'", May 1987.
- 2. "SAIER Model for Evaluation or' Loas-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants", NEDE-30996-1, June 1986.
- 3. "GE Reload Fuel Applicat. ton for Oyster Creek", NED0-24195, (As Amended).
- 4. "General Electric Standard Application for Reactor Fuel,"
NEDE-24011-P-A-8, May 1986.
)
! i 7-1/7-2
APPENDIX A l OYSTER CREEK SYSTEM RESPONSE CURVES i
I j
a l I i
l
NED0-31462 i
i 1
APPENDIX A l OYSTER CREEK SYSTEM RESPONSE CURVES l This appendix contains the system response curves for Oyster Creek.
Table A-1 contains the figure numbering sequence for the recirculation line breaks, and Table A-2 contains the figure numbering sequence for the non-recirculation line breaks.
i 1
I A-1 l
. _ --____ . . _ _ _ . - _ _ _ - _ _ - . _ _______m - _ _
j Table A-1 OiSTER CREEK RECIRC'ULATION LINE BREAK FIGURE
SUMMARY
DBA DBA DBA 80% DBA 80% DBA 60% DBA 60% DBA 40% DBA 40% DBA (NOM)
(App K) (NOM) (NOM) (App K) (NOM) (App K) (NOM) (App K)
Break DSCG DSCG Suction DSCG DSCG DSCG DSCG DSCG DSCG Failure Hot and Average la 2a 3a 4a 5a 6a 7a 8a 9a Channel Water Level Reactor Vessel Ib 2b 3b 4b Sb 6b 7b 8b 9b Pressure ,
g o
T Peak Cladding Ic 2c 3c 4c Sc 6e 7c 8e 9e
- Temperature h g
n Hot Channel Id 2d Od 4d 5d 6d 7d 8d 9d Heat Transfer Coefficioat j
9
- . y-,. , ,,, , . . . _
, , , . -- 't - - ~ ~ - ' " '- *' " '- - - - - - - - - - - - - - - - -
Table A-1 (Continued) 1.0 ft2 0.5 ft2 0.1 ft2 0,05 fg2 DBA DBA (Nom) (Non} (Nom) (Nom) (Nom) (App K)
DSCG DSCG DSCG DSCG DSCC Break DSCG l
l Failure l
Hot and Average 10a lla 12a 13a 14a 15a Channel Water Level Reactor Vessel 10b lib 12b 13b 14b 15b z
, Pressure hl I
Peak Cladding 1.0c lic 12c 13c 14c 15c M n
e Temperature Hot Channel Heat 10d lid 12d 13d 14d 1.5 '
! Transfer Coefficient l'
Oxide Thickness - - -
NEDO-31463 Table A-2 OYSTER CREEK NON-RECIRCULATION LINE BREAX FIGURE SLHMARY l
1*
Steamline Steaaline j (Inside (Outside Feedwater 1
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Hot and 36a 17a 18a 29a '
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