ML20237D829
| ML20237D829 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/30/1987 |
| From: | Forbes H, Wei P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML17055D453 | List: |
| References | |
| NEDO-31446, NUDOCS 8712240222 | |
| Download: ML20237D829 (116) | |
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NINE MILE POINT UNIT ONE SAFERICORECOOL/GESTR LOCA-LOSS-OF COOLANT ACCIDENT ANALYSIS P. W El H.R.FORBES a -,,, p ;m o P E'; ' B,Qi '),[,
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NED0-31446 Class I June 1987 l
NINE MILE POINT NUCLEAR GENERATING STATION UNIT ONE SAFER /CORECOOL/GESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS P. Wei H. R. Forbes Approved:
L. D. Noble, Manager Relo NLclear' Engineer-1 Approved:
M R. gAag, Manager
_Inttensing & Consulting Services NUCLEAR ENERGY BUSINESS OPERATIONS. GENERAL ELECTRIC COMPAN)
SAN JOSE, CALIFORNIA 95125 GENERAL $ ELECTRIC
NED0-31446.
CLASS I IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only o;;isttak'.ngs of the General Electric Company respecting informa-tion in this document are contained in the contract between NMPC 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 NMPC 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
'NED0-31446 CLASS I CONTENTS Page
1.0 INTRODUCTION
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 COREC00L 2-2 3.0 ANALYSIS PROCEDURE 3-1 3.1 BWR/2 Generic Analysis 3-1 3.2 Nine Mile Point-1 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 Break 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 Three-and Four-Loop Operation 5-3 5.3.2 Reduced Core Flow Operation (ELLLA) 5-4
6.0 CONCLUSION
S 6-1
7.0 REFERENCES
7_y APPENDIX - SYSTEM RESPONSE CURVES A-1 111/1v
NED0-31446 CLASS I LIST OF TABLFS Table Title Page 3-1 Analysis Assumptions for Nine Mile Point-1 Calculations 3-3 4-1 Operational and ECCS Parameters 4-2 4-2 Single-Failure Evaluation for Nine Mile Point-1 4-4 5-1 Summary of Recirculation Line Break Results - 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 MLPLHGR vs. Average Planar Exposure, Five-Loop Operation, 5-8 P6DRB299 Fuel 5-Sa Summary of Four-Loop and Three-Loop MAPLHGR Multipliers 5-9 Evaluation 5-Sb MAPLHCR Multipliers for P8DRB299 Fuel 5-10 A-1 Recirculation Line Break Figure Summary A-2 A-2 Non-Recirculation Line Break Figure Summary A-4 v/vi
NED0-31446 CLASS I 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 5-11 A-1 DBA DSCG (Nominal)
Water Level in Hot and Average Channel A-5 a
b Reactor Vessel Pressure A-6 c
Peak Cladding Temperature A-7 d
Heat Transfer Coefficient A-8 A-2 DBA DSCG (Appendiz K)
Water Level in Hot and Average Channel A-9 a
b Reactor Vessel Pressure A-10 c
Peak Cladding Temperature A-11 d
Heat Transfer Coefficient A-12 A-3 DBA Suction (Nominal)
Water Level in Hot and Average Channel A-13 a
b Reactor Vessel Pressure A-14 c
Peak Cladding Temperature A-15 d
Heat Transfer Coefficient A-16 A-4 80% DBA DSCG (Nominal)
Water Level in Hot and Average Channel a
A-17 b
Reactor Vessel Pressure A-18 c
Peak Cladding Temperature A-19 d
Heat Transfer Coefficient A-20 A-5 80% DBA DSCG (Appendiz K)
Water Level in Hot and Average Channel A-21 a
b Reactor Vessel Pressure A-22 Peak Cladding Temperature c
A-23 d
Heat Transfer Coefficient A-24 A-6 60% DBA DSCG (Nominal)
Water Level in Hot and Average Channel a
A-25 b
Reactor Vessel Pressure A-26 c
Peak Cladding Temperature A-27 d
Heat Transfer Coefficient A-28 vil
NED0-31446 CLASS.I LIST OF FIGURES (Continued)
Figure Title Page A-7 60% DBA DSCG (Appendix K) a Water Level in Hot and Average Channel A-29 I
b Reactor Vessel Pressure A-30 c
Peak Cladding Temperature A-31 d
Heat Transfer Coefficient A-32 l
l A-8 40% DBA DSCG (Nominal) a Water Level in Hot and Average Channel A-33 b
Reactor Vessel Pressure A-34 c
Peak Cladding Temperature A-35 d
Heat Transfer Coefficient A-36 A-9 40% DBA DSCG (Appendix K) a Water Level in Hot and Average Channel A-37 b
Reactor Vessel Pressure A-38 c
Peak Cladding Temperature A-39 d
Heat Transfer Coefficient A-40 2
A-10 1.0 Ft DSCG (Nominal) a Water Level in Hot and Average Channel A-41 l
b Reactor Vessel Pressure A-42 l
c Peak Cladding Temperature A-43 d
Heat Transfer Coefficient A-44 A-11 0.5 Ft2 DSCG (Ncainal) l a
Water Level in Hot and Average Channel A-45 b
Reactor Vessel Pressure A-46 c
Peak Cladding Temperature A-47 I
l d
Heat Transfer Coefficient A-48 A-12 0.1 Ft2 DSCG (Nominal) a Water Level in Hot and Average Channel A-49 I
b Reactor Vessel Pressure A-50 l
c Peak Cladding Temperature A-51 d
Heat Transfer Coefficient A-52 viii s
NEDO-31446 CLASS I LIST OF FIGURES (Continued)
Figure Title Page A-13 0.05 Ft2 DSCG (Nominal)
Water Level in Hot and Average Channel A-53 a
b Reactor Vessel Pressure A-54 c
Peak Cladding Temperature A-55 d
Heat Transfer Coefficient A-56 A-14 DBA DSCG - High Exposure (Nominal)
Water Level in Hot and Average Channel A-57 a
b Ree tor Vessel Pressure A-58 c
Peas Cladding Temperature A-59 d
Heat Transfer Coefficient A-60 A-15 DBA DSCG - High Exposure (Appendix K)
Water Level in Hot and Average Channel A-61 a
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)
Water Level in Hot and Average Channel A-66 a
b Reactor Vessel Pressure A-67 Peak Cladding Temperature A-68 c
d Heat Transfer Coefficient A-69 A-17 Steam Line Inside Containment (Nominal) kater Level in Hot and Average Channel A-70 a
b Reactor Vessel Pressure A-71 c
Peak Cladding Temperature A-72 d
Heat Transfer Coefficient A-73 A-18 Steam Line Outside Containment (Nominal)
Water Level in Hot and Average Channel A-74 a
b Reactor Vessel Pressure A-75 c
Peak Cladding Temperature A-76 d
Heat Transfer Coefficient A-77 1x
1 NED0-31446 CLASS I-LIST OF FIGURES (Continued)
Figure Title Page A-19 Feedwater Line (Nominal) a Water Level in Hot and Average Channel A-78 b
Reactor Vessel Pressure A-79 A-80 c
Peak Cladding Temperature d
Heat Transfer Coefficient A-81 1
x
NED0-31446 CLASS I
1.0 INTRODUCTION
The purpose of this document is to provide the results of the loss-of-coolant accident (LOCA) analysis for the Nine Mile Point Nuclear Power Station Unit 1.
The analysis was performed using the 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 conformance with the ECCS acceptance cri-teria of 10CFR50.46. The objective of the LOCA analysis contained hercin is to provide assurance that the most limiting break 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 Nine Mile Point Unit i values of the peak cladding tempera-ture (PCT) and maximum oxidation fraction for use in licensing evaluations are calculated for the limiting break.
The results conform to all the require-ments of 10CFR50.46 and Appendiz K.
The methodology described in this report will serve as the evaluation basis for future Nine Mile Point-1 fuel designs.
1-1/1-2
NEDO-31446 CLASS I i
2.0 DESCRIPTION
OF MODEL The General Electric evaluation model used for the Nine Mile Point Unit 1 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 after the core is uncovered and performs 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 calculate 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 for 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 such as heat transfer coefficients and PCT as a function of time.
SAFER also provides initial and boundary con-ditions, for the high power fuel bundle, to COREC00L.
2.1.2 GESTR-LOCA The GESTR-LOCA 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 SAFER.
GESTR-LOCA also initialized the transient pellet-cladding gap conduc-tance in SAFER.
2-1
NED0-31446 CLASS I 2.1.3 COREC00L COREC00L is a model for evaluation of core heatup transients for a fuel 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.,
21700*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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NED0-31446 CLASS I 3.0 ANALYSIS PROCEDURE 3.1 BWR/2 GENERIC 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 combination 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 for 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 demonstrate that a discharge coefficient of in the Moody Slip Flow Model l
yields the highest calculated PCT.
Comparison of the Appendiz K licensing basis and the upper bound (95th percentile) results demonstrated the conservatism of the BWR/2 licensing application methodology.
3.2 NINE MILE POINT-1 SPECIFIC ANALYSIS 3.2.1 Brask Spectrum Evaluation The plant-specific analysis performed for Nine Mile Point-1 consisted of break sizes ranging from 0.05 ft2 to the maximum of a DBA recirculation line 2
break (<5 45 ft ).
This plant-specific analysis evaluated recirculation line and non-recirculation line breaks, as well as an assessment of limiting break location and ECCS component failure.
The analysis assumptions (nominal and Appendix K) are presented in Table 3-1.
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
- GE Proprietary Information has been deleted.
3-1
NED0-31446 CLASS I nominal case. The limiting scenario for a spectrum of break sizes was then analyzed again with specifications for the Appendix K calculation (see Table 3-1).
The results of the Nine Mile Point-1 nominal and Appendix K cases were compared to assure that the PCT trends as a function of break size are 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 BWR/2 LOCA annlysis require that the cal-culated PCT following a postulated LOCA shall not exceed 2200*F and that the calculated maximum cladding local oxidation fraction shall not exceed 17%.
For a BWR/2 plant the ECCS performance is limited by different factors as the fuel exposure increases.
3-2
NED0-31446 CLASS I Table 3-1 ANALYSIS ASSUMPTIONS FOR NINE MILE POINT-1 CALCULATIONS Nominal Appendix K 1.
Decay Heat 1979 ANS 1971 ANS + 20%
(see Figure 3-1) 2.
Transition Boiling Iloeje Correlation 300*F Temperature 3.
Break Flow 1.25 EEM (SUB)
Moody Slip EEM (SAT) 4.
Metal-Water Reaction Cathcart Baker-Just 5.
Core Power 100%
102%
6.
MAPLHGRa (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 l
Bundle (2 Core Sprays) 10.
Fuel Stored Energy Best-Estimate GESTR Best-Estimate GESTR 11.
Rod Internal Pressure Best-Estimate GESTR Best-Estimate GESTR 12.
Cladding Rupture Stress BWR Design Values BWR Design Values abased on PLHGR of (low exposure) or (high exposure).
A multiplier of 1.02 was applied to the Appendix K values.
3-3
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NED0-31446 CLASS I 4.0 INPUT TO ANALYSIS 4.1 PLANT-SPECIFIC PARAMETERS A list of the significant Nine Mile Point-1 plant-specific input param-eters to the LOCA analysis is presented in Table 4-1.
Table 4-2 identifies the break locations and corresponding single-failure / system available com-binations specifically evaluated for Nine Mile Point-1.
4.2 TIMING FOR THE ONSET OF BOILING TRANSITION The current Nine Mile Point-1 LOCA lictnsing report (Reference 3) con-cludes, 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 DBA), where there is very rapid flow coastdown, the duration of nucle-to ate boiling following the break is calculated using the GE dryout correlation (in the CHASTE code)', which is based on instantaneous flow stagnation condi-tions.
(The LAMB, SCAT and CHASTE models and applications are described in Reference 4.)
In this Nine Mile Point-1 SAFER Analysis, for break sizes from DBA to the break spectrum evaluation (results summarized in Section 5.0) l 3
utilized a timing for the onset of boiling transition based on the GE dryout correlation. For break sizes smaller than
, nucleate boiling was assumed to be maintained until core uncovery.
This approach established that the Nine Mile Point-1 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-nario and the resultant MAPLHGR calculations.
4-1 l
.NEDO-31446 CLASS I Table 4-1 OPERATIONAL AND ECCS PARAMETERS A.
Plant Parameters Core Thermal Power (MWth)
Nominal 1850 (100% of Rated)
Appendix K-18B7 (102% of Rated)
Vessel Steam Output (Iba/hr)
Vessel Steam Dome Pressure (psia)
Maximus Recirculation Line 5.446 2
Break Area (ft )
Initi.1 MCPR Initial Water Level Scram Trip Level B.
Emergency Core Cooling System Parameters Core Spray' System l
Vessel Pressure at Which Pump Can Deliver Flow (psid)
Systen Flow at Vessel Pressure (psid) for One Loop (gpa)
Initiating Signals and Setpoints Low Water Level or j
I High Drywell Pressure (psig)
Runout Flow (at Zero paid) for Each Loop Maximus Allowable Delay Time from Initiating Signal to Pump at Rated Speed (sec) l Injection Valve Stroke Time (sec)
Pressure Permissive at Which Injection Valve Opens (psid) 4 Core Spray Flow to Hot Bundle (2 headers) l (gpa)
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NEDO-31446 CLASS I Table 4-1 (Continued)
OPERATIONAL AND ECCS PARAMETERS i
ADS Total Number of Valves in System Number of Valves Assumed in Analysis Minimus Flow Capacity of 3 Valves (1ba/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 Initiating Signals-Low Water Level or High Vessel Pressure (psia)
Maximum Isolation Valve Stroke Time (sec)
Maximum Operating Pressure (Vessel) (psia)
Initial Operating Temperature on Shell Side of Condenser (*F)
Initial Water Mass on Shell Side of Each condenser (Gallons)
Surface Heat Transfer Area of Each Condenser (ft2) l 4-3
NED0-31446 CLASS I Table 4-2 SINGLE-FAILURE EVALUATION FOR NINE MILE POINT-1
'Brsak Location Single Failure Available Systema R: circulation Line Fs2dwater and Main Stemalinea Ctre Spray Line EC = Emergency Condenser CS = Core Spray ADS = Automatic Depressurization System 4-4 t.
NED0-31446 CLASS I 5.0 PLANT-SPECIFIC RESULTS 5.1 BREAK SPECTRUM CALCULATIONS 5.1.1 R_ecirculstion 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 Appen-dix 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 s
peak cladding temperature of (low exposure). The corresponding PCT for this break with Appendix K specified models was calculated to be and for the respectively. Plots showing system i
responses for all break spectrue cases are presented in the Appendix to this I
report.
5.1.2 Non-Recirculation Line Breaks Evaluations were also performed for some of the non-recirculation line breaks. These breaks (including feedwater, core spray and main steam lines)
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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 sustain essentially no heatup and are far from becoming candidates for the limiting event. The same conclusion applies for break sizes smaller than i
the guillotine break for these lines. The system responses of these breaks are also presented in the Appendix.
5-1 i
NED0-31446 CLASS I Nine Mlle Point-1 plant-specific evaluations were not performed for other non-recirculation line breaks (e.g., EC lines, liquid instrument lines, cleanup system lines, etc.).
Ibese non-recirculation line breaks will not become can-I didates for the limiting event, since they are essentially the same as small recirculation or steam line breaks.
l 5.2 TECHNICAL SPECIFICATION MAPLHGR LIMITS GE BWR MAPLHGR limits (as a function of fuel exposure) are based on the l
For BWR/2 plants, in general, and the Nine Mile Point-1 plant, spec-ifically, the MAPLHGR calculated from the is limiting and datermines the Technical Specification limits.
The MAPLHGR limits for the P8DRB299 fuel bundle were evaluated as a fune-tion of exposure with the limiting scenario identified in the break spectrum analyses.
Table 5-4 lists the P8DRB299 MAPLHGR limits (five-loop operation) along with the calculated PCT and peak local oxidation fraction. The highest PCT calculated was at an exposure of
, while the highest peak i
i local oxidation fraction calculated was Based on the results for the P8DRB299 bundle, it is erpected that the SAFER evaluation methodology would show that the current MAPLHGR limits for fusi types 8DNB277 and PP.DNB277 (calculated by Reference 3) are conservative at all exposures. Therefore, no change will be made in the MAPLHGRs for these two fuel types because they are not expected to limit core operation.
l 5-2
NED0-31446 CIASS 1 For fuel types in future Nine Mile Point-1 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 typ9 specific MAPLHGRs.
5.3 ALTERNATE OPERATING MODE CONSIDERATIONS
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5.3.1 Three-and Four-Loop Operation The Reference 3 analysis of three-and four-recirculation-loop operation identified two main differences in the LOCA analysis as compared to the normal five-loop case:
l l
l The effects of these differences on the SAFER calculations will depend on the break size.
5-3
NED0-31446 i
CLASS I Evaluation results for four-loop and three-loop operation, with the inoperative loops isolated, are summarized in Table 5-Sa.
The DBA break (low 2
cod high exposure) and the 0.1 ft break (highest small-break PCT from the nominal analysis) cases were calculated with a compared with_the five-loop base cases. The results show that a _
is adequate to compensate for the effect of loss of inventory and the faster coastdown for the large and small breaks. Table 5-5b summarizes the four-loop and three-loop MAPLHGR multipliers for the P8DRB299 fuel.
5.3.2 Reduced Core Flow Operation (ELLLA) i The impact, on MAPLHGR limits, of operating at rated reactor power and reduced core flow (i.e., in the Extended Load Line Limit Analysis (ET.T.TA) f 5-4
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. CLASS I Table 5-1
SUMMARY
OF RECIRCULATION LINE BREAK RESULTS -
NOMINAL EVALUATION (l>
' Core-Wide
. Break PCT Peak Local Metal-Water Size (ft2)
(*F)
Oxidation (%)
Reaction (%)
Discharse Break - Low Exposure (14 GWd/MTU)
Discharse Break - High Exposure (22 GWd/MTU) l Suction Break - Low Exposure (14 GWd/MTU)
(2)Less than Appendir K low exposure DBA.
5-5
l NEDO-31446 l'
CIASS I -
Table 5-2
SUMMARY
OF RECIRCULATION LINE BREAK RESULTS -
APPENDIX K EVALUATION (1)
Core-Wide Break 2 PCT Peak Local Metal-Water Size (ft )
(*F)
Oxidation (%)
Reaction (%)
Diccharge Break - Low Exposure (14 GWd/MTU)
Dircharge Break - High Exposure (22 GWd/MTU)
(2)Less than Appendir K low exposure DBA.
l 5-6 I
NEDO-31446 CLASS I Table 5-3 i
SUMMARY
OF NON-RECIRCULATION LINE REAK RESULTS -
NOMINAL EVALUATION (1 Core Wide PCT Peak Local Metal-Water
(*F)
Oxidation (%)
Reaction (%)
Core Spray Line break Steamline Break Inside Containment Steam 11ne Break Outside Containment Feedwater Line Break (2)Less than Appendix K low exposure DBA.
5-7
-NEDO-31446 CLASS I Table 5-4 MAPLHGR vs.' AVERAGE PIANAR EXPOSURE FIVE-LOOP OPERATION Plant: NMP-1 Fuel Type: P8DRB299 Average Planar Exposure MAPLHGR PCT Local (GWd/MTU)
(kW/ft)
(*F)
Oxidation Fraction 0.22 10.9 1.1 10.9 5.5 10.9 11.0 10.9 14.0 10.8 15.0 10.7 16.5 10.4 22.0 9.7 27.5 9.6 33.0 9.5 38.5 9.3 44.0 9.2 50.0 9.0 0
1 5-8 l
NED0-31446 CLASS I Table 5-Sa
SUMMARY
OF FOUR-LOOP AND THREE-LOOP MAPLEGR MULTIPLIERS EVALUATION P8DRB299 FUEL, ISOLATED CONDITION Loops Parameter Five*
Four*
Three**
MAPLHGR Multiplier DBA Break PCT (*F)
Low Exposure High Exposure /0xide Thickness (%)
2 Small (0.1 ft ) Break PCT (*F)
Notes
- Four-loop and five-loop are evaluated with identical Appendix K conditions.
5-9
1 i
i NEDO-31446 CLASS I Table 5-Sb MAPLHGR MULTIPLIERS FOR P8DRB299 FUEL t
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l CResults are applicable to all exposure ranges.
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NED0-31446 CLASS I
6.0 CONCLUSION
S The recirculation line break results presented in Section 5.1 demonstrate that a sufficient number of Nine Mile Point-1 plant-specific PCT points have been evaluated te verify that the trend of the PCT curves, for both the nominal and Appendix K calculations, is similar to the (Reference 2) generic PCT versus break size curves.
It has been demonstrated generically that the PCT calculated in accor-dance 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 condi-tions. These calculations were performed to qualify the " Appendix K Proce-dure" as being sufficiently conservative. The generic upper bound PCTs, which includes a 50*F conservatism (AS) assigned by the NRC (Reference 1), were and for low and high exposures, respectively.
By 'omparison, l
c the generic licensing basis Appendix K evaluation with SAFER /COREC00L for the limiting conditions provided PCTs of and for low and high expo-sures, providing (low exposure) and (high exposure) margins to the upper bound requirements.
The Nine Mile Point-1 plant-specific Appendix K analysis will have simi-lar margin to the 95th percentile PCT because of the following considerations:
(1) Nine Mile Point-1 has an ECCS configuration identical to that used in the generic BWR/2 analysis. Therefore, the limiting case LOCA for both Nine Mile Point-1 and the generic BWR/2 is the (2) The key operating parameters for the plant-specific Nine Mile Point Unit 1 analysis are similar to the inputs used in the calculations of the generic analysis PCT.
6-1
NEDO-31446 CLASS I (3).The similarity between the generic and the plant-specific evalua-tions (in plant configuration and the operating parameters) is responsible for the similar PCTs calculated with SAFER /COREC00L.
The generic nominal analysis reported and (for the limiting scanario), while the plant-specific analysis yielded and for the low and high exposures, respectively. The Appendiz K licensing basis results wara also very similar.
I Therefore, it is confirmed that the generic assessment (Reference 2) is cpplicable to Nine Mile Point-1 and the Nine Mile Point-1 Appendix K licens-I ing basis analysis exceeds the upper bound 95th percentile PCT. Also, the picnt-specific results of the Appendix K licensing analysis of Section 5.0 seat the criteria of 10CFR50.46.
In conclusion, it is verified that the Nine Mile Point-1 plant-specific l
SAFER /CORECOOL/GESTR-LOCA analysis meets the explicit requirements of the Rafarence 1 NRC Safety Evaluation Report.
I 6-2
1
).
NED0-31446
[
CLASS I
7.0 REFERENCES
1.
Letter, A. C. Thadani (NRC) to H. C. Pfefferien (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.
" SAFER Model for Evaluation of Loss-of-Coolent Accidents for Jet Pump and Non-Jet Pump Plants", NEDE-30996-P, June 1986.
3.
"LOCA Analysis Report for Nine Mile Point Unit 1 NGS", NEDO-24348, August 1981.
4.
" General Electric Standard Application for Reactor Fuel (U.S. Supplement)",
NEDE-24011-P-A-8-US, May 1986.
l 7-1/7-2
l NED0-31446 CLASS I l
l APPENDIX NINE MILE POINT 1 SYSTEM RESPONSE CURVES
1 NEDO-31446 CLASS I APPENDIX NINE MILE POINT 1 SYSTEM RESPONSE CURVES This appendix contains the system response curves for Nine Mile Point 1.
Table A-1 contains the figure numbering sequence for the recirculation line breaks, and Table A-2 contains the figure nukhering sequence for the non-k recirculation line breaks.
l l
A-1
D G
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A B)
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SUMMARY
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Line Failure Hot and 16a 17a 18a 19a Avarage Ch nnel Water Level
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