ML20128A028
| ML20128A028 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 06/30/1985 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20128A015 | List: |
| References | |
| NUDOCS 8507020545 | |
| Download: ML20128A028 (65) | |
Text
,
PNPP MAXIMUM EXTENDED OPERATING DOMAIN ANALYSIS June 1985 l
Prepared for Cleveland Electric Illuminating Company ,
Perry 1 & 2 Nuclear Power Stations Prepared by General Electric Company Nuclear Energy Business Operations San Jose, California 95125 0507020545 850627 PDR ADOCK 05000440 A PDR
APPENDIX 15.E TABLE OF CONTENTS PAGE 15.E Maximum Extended Operating Domain 15.E.1-1
! 15.E.1 Definition of Current Power / Flow Operating Domain 15.E.1-1 15.E.2 Introduction and Summary 15.E.2-1 15 E.3 MCPR Operating Limit 15.E.3-1 15.E.3.1 Abnormal Operating Transient 15.E.3-1 15.E.3.2 Rod Withdrawal Error 15.E.3-2 I
15.E.4 Fuel Integrity - Stability 15.E.4-1 15.E.5 Loss of Coolant Accident Analysis 15.E.5-1 15.E.6 Containment Response Analysis 15.E.6-1 l
15.E.7 Loads Impact on Internal 15.E.7-1 15.E.7.1 Acoustic and Flow Induced Loads 15.E.7-1 '
15.E.7.2 Reactor Internal Pressure Differences 15.E.7-1 15.G.7.3 Impact on Reactor Internals 15.E.7-1 15.E.8 Flow Induced Vibration 15.E.8-1 15.E.9 Impact on Anticipated Transient Without Scram 15.E.9-1 l
15.E.10 Fuel Mechanical Performance 15.E.10-1 15.E.11 Partial Feedwater Heating Operation in the Extended Operating Domain 15.E.11-1 15.E.11.1 Abnormal Operating Transients 15.E.11-2 15.E.11.2 Other Evaluations, PFH Operation in ME0D 15.E.11-11 15.E.12 References 15.E.12-1 O
j 15.E.i
< DTS:pc:rf/LO4034*-2 6/12/85
4 LIST OF TABLES i
NUMBER TITLE PAGE i l
l 15.E.2-1 Summary of Rated Operating Limit MCPR Values 15.E.2-3 l 15.E.3-1 Input Parameters and Input Conditions for Transient Analysis for MEOD at 104.2% Power /
75% Core Flow 15.E.3-3,4
, 15.E.3-2 Input Parameters and Input Conditions for Transient Analysis for ME0D at 104.2% Power /
1 105% Core Flow 15.E.3-5,6 15.E.3-3 Summary of Transient Peak Values Results -
ME0D Operation 15.E.3-7 l 15.E.3-4 Summary of Critical Power Ratio Results -
ME0D Operation 15-E.3-8 15.E.3-5 Analysis Power - Flow Points for Transient :
Evaluation 15.E.3-9 15.E.11-1 Summary of Transient Peak Values Results -
PFH in ME0D, Beyond End of Cycle; 250 F 15.E.11-3 15.E.11-2 Summary of Transient Peak Values Results -
PFH in ME00 Beyond, End of Cycle; 370 F, 320 F 15.E.11-4,5 15.E.11-3 Summary of Transient Peak Values Results -
PFH in ME00, 2000 MWD /T B*efore End of Cycle 15.E.11-6 15.E.11-4 Summary of Critical Power Ratio Results -
PFH in MEOD, Beyond End of Cycle; 250"F 15.E.11-7 15.E.11-5 Summary of Critical Power Ratio Results -
PFH in ME00, Beyond End of Cycle; 370'F, 320'F 15 E.11-8,9 15.E.11-6 Summary of Critical Power Ratio Results -
PFH in ME00, 2000 MWD /T Before End of Cycle 15.E.10 15.E-fi 4 i
DTS:pc:rf/LO4034*-3 6/12/85
LIST OF FIGURES NUMBER TITLE 15.E.2-1 Maximum Extended Operating Domain Power / Flow Map (MEOD) 15.E.3-1 Load Rejection With Bypass Failure 104.2% Power /
105% Flow 15.E.3-2 Load Rejection With Bypass Failure 104.2% Power /
73.6% Flow 15.E.3-3 Feedwater Controller Failure 104.2% Power /105% Flow 15.E.3-4 Feedwater Controller Failure 104.2% Power /
73.6% Flow 15.E.11-1 Load Rejection With Bypass 104.2% Power /
105% Flow 250 F rated feedwater temperature 15.E.11-2 Feedwater Controller failure 104.2% Power /
105% Flow 250 F rated feedwater temperature 15.E.11-3 Load Rejection With Bypass 104.2% Power /
108.7% Flow 370"F rated feedwater temperature 15.E.11-4 Load Rejection Without Bypass 104.2% Power /
110.0% Flow 320 F rated feedwater temperature 15.E.11-5 Feedwater Controller Failure 104.2% Power /
108.7% Flow 370*F rated feedwater temperature 15.E.11-6 Feedwater Controller Failure 104.2% Power /
110% Flow 320*F rated feedwater temperature 15.E.11-7 Load Rejection Without Bypass 104.2% Power /
74.8% Flow 370'F rated feedwater temperature 15.E.11-8 Load Rejection Without Bypass 104.2% Power /
73.7% Flow 320'F rated feedwater temperature
- 15.E.11-9 Feedwater Controller Failure 104.2% Power /
74.8% Flow 370'F rated feedwater temperature l 15.E.11-10 Feedwater Controller Failure 104.2% Power /
73.7% Flow 320*F rated feedwater temperature 1
15.E-iii DTS:pc: rf/LO4034*-4 6/12/85
15.E Maximum Extended Operating Domain This appendix provides the justification that the operating domain as shown in Figure 4.4-2 can be extended (See Figure 15.E.2-1) and still meets all the requiremeats established by the Code of Federal Regulations with PNPP's 100%
full power license conditions.
15.E.1 Definition of the Current and the Maximum Extended Power / Flow Operating Domains The current power / flow operating domain as given in Figure 4.4.2 of Chapter 4 can be regarded as a map bounded by the following restrictions:
(1) The 100% rated power limit.
(2) The 105% rated steam flow rod line.
(3) The 100% rated core flow condition.
(4) Low power recirculation system component cavitation restriction.
(5) Minimum core flow resultant from restrictions on pump speed and FCV position.
The Maximum Extended Operating Domain is essentially extending additional operational power / flow areas to the operating domain given in Figure 4.4-2.
They are:
(a) The Extended Load Line Region (ELLR) - the area where higher power can be achieved at lower than rated core flow conditions.
(b) The Increased Core Flow Region (ICFR) - the area where core flow up to 105% rated is utilized.
DTS:ge: rf/F08013* 15.E.1-1 6/12/85
15.E.2 Introduction & Summary This appendix presents the results of a safety and impact evaluation for i operation of the Perry Nuclear Power Plants (PNPP) in an expanded operating ,
envelope called the Maximum Extended Operating Domain to permit improved power ascension capability to full power as well as to provide additional flow range at rated power including an increased flow region to compensate for reactivity reduction due to exposure during an operating cycle.
The total Maximum Extended Operating Domain (ME00) is shown in Figure 15.E.2-1.
The extended load line region (ELLR) boundary is limited by 75% core flow at 100% power and its corresponding power / flow constant rod line. This is deter-mined based on a safety and impact evaluation in meeting thermal and reactivity margins. The Increased Core Flow Region (ICFR) is bounded by the 105% core j flow line. This ICFR boundary is limited by plant recirculation system capa-l bility, acceptable flow induced vibration and force impact on the vessel internal components.
l The ME00 evaluation is performed for PNPP on a 18-month equilibrium cycle basis and is applicable to 12-month or 18-month cycle operation for both initial and j reload cycles with the current GE 6 fuel design. The results of this evaluation i are:
(a) The limiting normal and abnormal operating transients in Chapter 15 were reevaluated for the ME0D conditions. It is also determined that the fuel mechanical limits are met for all transients occurring in the ME00.
1 (b) The Loss of Coolant Accident and Containment responses as described in Chapter 6 were reevaluated in the ME00. It is found that the responses are bounded by the current design analysis.
(c) Thermal hydraulic stability was evaluated for its adequacy with respect to j the General Design Criterion 12 (10CFR50, Appendix A). It is shown that ME0D operation satisfies this stability criterion.
l DTS: ge: rf/F08013* 15.E.2-1 6/12/85 L___________.----_. - - - - _ - - _ _ _ _ _ . - - - - ___ ___ _ _ _ _ _ _ _ _ _ _ _ _ . - - - - - - - - - - - - _ - - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
(d) The effect of increased flow induced loads due to increased core flow on l the reactor internal components and fuel channels are assured for their design adequacy. The effect of increased flow rate on the flow induced vibration response of the reactor internal will be monitored during startup testing and be evaluated to ensure the responses are within acceptable limits for PNPP.
(e) Several impact evaluations were also performed to justify operation in the ME0D. It was found that acceptance criteria and design limits are met.
(f) This appendix also justifies Partial Feedwater Heating (PFH) operation described in Section 150 for rated feedwater temperature ranging from 420 F to 320 F during and beyond the operating cycle in the ME0D (ELLR and ICFR), and rated feedwater temperature ranging from 320 F to 250 F beyond the end of cycle in the ICFR. All evaluations described in Appendix 150 were reevaluated or reviewed in the ME0D to ensure that PFH operation in this maximum extended operation region is safe and feasible with the required additional modifications to the Technical Spec,ification MCPR limits.
(g) A summary of the rated operating limit MCPR value for various modes of
- operation is tabulated in Table 15.E.2-1.
Even though the MEOD boundary is set at 105% rated core flow in the ICFR, the transient analyses covered in this appendix include core flow as high as 116%
rated (see Table 15.E.3-5). It is customary during the reactor startup test program to test the recirculation flow at the full-open position of the recirculation flow control valve as long as no operating limit is exceeded.
For Perry, this high flow rate is estimated to be about 108% at 100% power and 112% at 56% power.
DTS:ge: rm/F08013* 15.E.2-2 6/25/85
r c-Table 15.E.2-1 Summary of Rated Operating Limit MCPR Values l
Mode of Operation Rated i OLMCPR l
Original FSAR Power Flow Map (Figure 4.4.2) 1.18 !
PFH (420 F to 370 F rated FWT) 1.18 PFH (370 F to 320 F rated FWT) 1.19 PFH* (320 F to 250 F rated FWT) 1.21 ME00 Power Flow Map (Figure 15.E.2-1) 1.18 PFH (420*F to 320*F rated FWT) 1.19 PFH* in ICFR (320 F to 250 F rated FWT) 1.21
- For beyond end of cycle only.
(1) All OLMCPRs are for initial core only. 0.01 needs to be added for reload application (2) All evaluations and results are for GE6 fuel for PNPP with E0C target Haling exposure distribution (3) Nomenclature:
}
PFH = Partial Feedwater Heating operation to be applied both during the operating cycle and beyond the end of cycle (see Appendix 150) 9 DTS:gc: rf/F08013* 15.E.2-3 l
6/12/85
e A. NATURAL CIRCULATION B. LOW RECIRC. PUMP SPEED VALVE MINIMUM POSITION (0% OPENING) l
.C. LOW RECIRC. PUMP SPEED VALVE MAXIMUM POSITION _
{ D. RATED RECIRC. PUMP SPEED VALVE MINIMUM POSITION (0% OPENING)
E. ANALYTICAL LOWER LIMIT-OF AUTOMATIC LOAD FOLLO: JING REGION F. RATED POWER FLOW -
G. LOWEST ALLOWABLE FLOW AT RATED POWER (100% P, 75%F) ,
_H. HIGHEST ALLOWABLE FLOW AT RATED POWER (100% P, 105% F) -
l KAGFK: EXTENDED LOAD LINE REGION FHIJF: INCREASED CORE FLOW REGION '
~
120 -
Increased Core 110
~
{
Flow Region (ICFR) j G F H 100 - ~
Extended Load Line Region
's s'
(ELLR) g E
w 90 . y 's Y
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N o
=
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o 80 -
4S y'U g N
y g
e a s, 100% ROD LINE N
- D s N '
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. 1
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20
- CAVITATION REGION ,
10 -
l
, , , , , , . . i e '
10 20 30 40 50 60 70 80 90 $00 110 120 CORE FLOW (%)
CLEVELAND ELECTRIC . 1 E 2-1 l
l l
l
15.E.3 MCPR Operatina Limit 15.E.3.1 Abnormal Operatina Transients All abnormal operating transients described in Chapter 15 were examined for Maximum Extended Operating Domain (ME00) operation. Three Limiting Abnormal Operating Transients are discussed here in detail. They are:
1 (a) Generator Load Rejection with Bypass Failure (LRNBP) l (b) Feedwater Flow Controller Failure (FWCF)
(c) 100'F-Loss'of Feedwater Heating The reevaluations were performed at various ME0D bounding power flow conditions of Figure 15.'E.2-1 at the end of the 18 month equilibrium cycle. Plant heat balance, core coolant hydraulics and nuclear transient parameter data were developed and used in the above transient ar.alysis. Full arc (FA) turbine control valve closure characteristics were assumed. The initial condition for the lowest and highest flow points at rated power are presented in Tables 15.E.3-1 and 15.E.3-2. The computer model described in ref. 15.E-12-1 was used to simulate both the Generator Load Rejection With Bypass Failure and Feedwater Controller Failure events. The transient peak values results and critical i power ratio (CPR) results for the two cases analyzed at 104.2% power (lowest and highest flow) are summarized in Table 15.E.3-3 and 15.E.3-4 respectively.
'The transient responses are presented in Figure 15.E.3-1 to 15.E.3-4. Several
( other power flow conditions on the MEOD boundary rod line and the maximum core -
l flow boundary were also analyzed. Table 15.E.3-5 identified these additional analyzed power flow points. The results of this evaluation show that the ACPR results for all the cases analyzed in the ME00 are bounded by the original
. Technical Specification limits.
The 100*F Loss of Feedwater Heating (LFWH) Transient results in Chapter 15 are f applicable to the ME0D.- A generic statistical LFWH analysis using the computer model described in Reference 15.E.12-2 and methodologies described in Reference 15.E.12-3 utilizes a large data base to establish bounding results DTS:gc:rf/F08013* 15.E.3-1 6/12/85
for this event. It is found that the LFWH response initiated from all MEOD off rated power / flow conditions is bounded by the rated condition bounding 95%
probability 95% confidence values. This generic bounding ACPR value is bounded by the value documented in Chapter 15 for this event.
Additionally, overpressure protection transient analysis using the computer model described in ref. 15.E.12-1 is performed at the various power flow conditions (Table 15.E.3-5). The bounding MSIV closure flux scram event resulted in a peak pressure of 1273 psig at a postulated 110% core flow condition. Therefore it is shown that the peak vessel pressure for the ME0D is below the ASME code limit of 1375 psig. Hence, adequate pressure margin is present for the Chapter 15 transients in the ME0D.
Furthermore, the pressure controller downscale failure event has been examined to show that establishing an operating limit MCPR value for this event is no longer necessary. The PNPP specific steam bypass failure (when the turbine control valves close) is only possible if there is a short in the cabling for the ground of the bypass and the test card. According to IEEE 500-1984
" Reliability for Nuclear Power Generating Stations", the probability for such unique and untimely failure is so remote that even an infrequent transient classification is too conservative.
15.E.3.2 Rod Withirawal Error The rod withdrawal error (RWE) transient documented in Chapter 15 Appendix B is analyzed using a statistical evaluation of the minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) response to the withdrawal of ganged control rods throughout the operating power / flow map including the ME0D region. Therefore, the current Technical Specification MCPR limit is adequate to protect the RWE in the MEOD.
DTS:ge:rm/F08013* 15.E.3-2 6/25/85
Table 15.E.3-1 Input Parameters and Initial Conditions for Transients and Accidents for MEOD, 104.2% Power, 73.6% Flow
- 1. Thermal Power Level, MWt 3729.3 (104.2% rated)
Analysis Value
- 2. Steam Flow, lb. per sec 4468 (104.4% rated)
Analysis Value
- 3. Core Flow, Ib per hr 76.5 x 106
- 5. Feedwater Temperature, F 425
- 6. Vessel dome pressure, psig 1044
- 7. Core exit pressure, psig 1052
- 8. Turbine Bypass Capacity, % NBR 35
- 9. Core Coolant Inlet Enthaply 520.2 Stu per lb
- 10. Turbine Inlet Pressure, psig 960
- 11. Fuel Lattice P8x8R
- 12. Core Leakage Flow, % 12.9
- 13. Required MCPR Operating Limit 1.27 First Core
- 14. MCPR Safety Limit for Incidents of Moderate Frequency First Core 1.06 Reload Cores -
1.07
- 15. Doppler Coefficient (-)C/ F 0.132 (a)
Analysis Data
- 16. Void Coefficient (-)C/% Rated Voids Analysis Data for Power Increase Events 14.0 (a)
Analysis Data for Power Decrease Events 4.0 (a)
DTS:ge:rf/F08013* 15.E.3-3 6/12/85
Table 15.E.3-1 (Continued)
- 17. Core Average Rated Void Fraction, % 48.7 18 Jet Pump Ratio, M 2.25
- 19. Safety / Relief Valve Capacity, % NBR 91210 psig 111.4 Manufacturer Dikker Quantity Installed 19
- 20. Relief Function Delay, seconds 0.4
- 21. Relief Function Response Time Constant, sec. 0.1
- 22. Analyses Inputs for Safety / Relief Valves Safety Function, psig 1175, 1185, 1195, 1205, 1215 Relief Function, psig 1145, 1155, 1165, 1175
- 23. Number of Valve Groupings Simulated Safety Function, No. 5 Relief Function, No. 4
- 24. High Flux Trip, % NBR Analysis Setpoint (122x1.042), % NBR 127.2
- 25. High Pressure Scram Setpoint, psig 1095
- 26. Vessel Level Trips, Feet Above Separator Skirt Bottom
- Level 8 - (L8), feet 5.89 Level 4 - (L4), feet 4.04 Level 3 - (L3), feet 2.165 Level 2 - (L2), feet (-) 1.739
- 27. APRM Thermal Trip Setpoint, % NBR 118.8
- 28. RPT Delay, seconds 0.14
- 29. RPT Inertia Time Constant for Analysis, sec. 5
- 30. Total Steamline Volume, ft3 3850 (a) These values for Reference 15.E12-4 analysis only. Reference 15.E12-1 values are calculated within the code.
DTS:gc:rm/F08013* 15.E.3-4 6/25/85
Table 15.E.3-2 Input Parameters and Initial Conditions for Transients and Accidents for ME0D,104.2% Power,105% Flow
- 1. Thermal Power Level, MWt 3729.2 (104.2% rated)
Analysis Value
- 2. Steam Flow, ib. per sec 4485 (104.8% rated)
Analysis Value
- 3. Core Flow, Ib per br. 109.2 x 106
- 5. Feedwater Temperature, F 425
- 6. Vessel dome pressure, psig 1045
- 7. Core exit pressure, psig 1057
- 8. Turbine Bypass Capacity, % NBR 35
- 9. Core Coolant Inlet Enthalpy 529.4 Btu per lb
- 10. Turbine Inlet Pressure, psig 960
- 11. Fuel Lattice P8x8R
- 12. Core Leakage Flow, % 12.9
- 13. Required MCPR Operating Limit 1.18 First Core
- 14. MCPR Safety Limit for Incidents of Moderate Frequency First Core 1.06 Reload Cores 1.07
- 15. Doppler Coefficient (-)C/ F 0.132(a)
Analysis Data -
9 DTS:gc:rf/F08013* 15.E.3-5 6/12/85
Table 15.E.3 (Continued) l
- 16. Void Coefficient (-)C/% Rated Voids Analysis Data for Power Increase Events 14.0 (a)
Analysis Data for Power Decrease Events 4.0 (a)
- 17. Core Average Rated Void Fraction, % 41.7
- 18. Jet Pump Ratio, M 2.25
- 19. Safety / Relief Valve Capacity, % NBR
@l210 psig 111.4 Manufacturer Dikker Quantity Installed 19
- 20. Relief Function Delay, seconds 0.4
- 21. Relief Function Response Time Constant, sec. 0.1
- 22. Analyses Inputs for Safety / Relief Valves Safety Function, psig 1175, 1185, 1195, 1205,1215 Relief Function, psig 1145, 1155, 1165, 1175
- 23. Number of Valve Groupings Simulated Safety Function, No. 5 Relief Function, No. 4
- 24. High Flux Trip, % NBR Analysis Setpoint (122x1.042), % NBR 127.2
- 25. High Pressure Scram Setpoint, psig 1095
- 26. Vessel Level Trips, Feet Above Separator Skirt Bottom Level 8 - (L8), feet 5.89 Level 4 - (L4), feet 4.04 Level 3 - (L3), feet 2.165 Level 2 - (L2), feet (-) 1.739
- 27. APRM Thermal Trip Setpoint, % NBR 118.8
- 28. RPT Delay, seconds 0.14
- 29. RPT Inertia Time Constant for Analysis, sec. 5
- 30. Total Steamline Volume, ft 3 3850 (a) These values for Reference 15.E.12-4 analysis only. Reference 15.E.12-1 values are calculated within the code.
DTS:gc:rm/F08013* 15.E.3-6 6/25/85
l Table 15.E.3-3 Summary of Transient Peak Values Results - 104.2% Power - MEOD(*)
Peak Peak Peak Peak Neutron Dome Vessel Steamline Core Flow Flux Pressure Pressure Pressure Transient (% NBR) (% NBR)(3) (psig) (psig) psig Figure Load Rejection 107.2(b) 268 1217 1251 1217 -
with Bypass Failure 105.0 259 1217 1251 1217 15E3-1 73.6 171 1220 1243 1222 15E3-2 Feedwater 107.2(b) 149 1172 1203 1172 -
Controller Failure, Max.
Demand 105.0 148 1172 1203 1171 15E3-3 73.6 111 1175 1195 1172 15E3-4 (a) Feedwater is 425 F.
(b) Maximum achievable core flow with 425*F feedwater.
15.E.3-7 DTS:pc/LO40411* i 4/19/85
l i
l l
l Table 15.E.3-4 Summary of CPR Results - 104.2% Power MEOD I ")
l Core Flow l Transient (% NBR) ICPR ACPR MCPR Load Rejection 107.2(b) 1.18 0.11 1.07 l
with Bypass Failure 105.0 1.18 0.11 1.07
[ 73.6 1.27 0.06 1.21 Feedwater 107.2(b) 1.18 0.09 1.09 l Controller Failure, Max.
I Demand l
105.0 1.18 0.09 1.09 73.6 1.27 0.09 1.18 l
l (a) Feedwater Temperature is 425*F.
(b) Maximum achievable core flow with 425 F Feedwater.
I 15.E.3-8 DTS:pc/LO40411*
4/19/85
-n .o.... . . . . .
Table 15.E.3-5 Analysis Power-Flow Points for Perry Bounding Transient Evaluation Power (%)/ Flow (%) Transients 70/40 LRNBP FWCF CLDLP FCVO 83/55 LRNBP 104.2/75 LRNBP FWCF FCVO 104.2/100 LRNBP FWCF 104.2/110 LRNBP FWCF 53.5/116 LRNBP FWCF NOTE:
LRNBP Generator Load Rejection With Bypass Failure FWCF Feedwater Controller Failure (maximum demand)
CLDLP Cold Loop Startup FCVO Flow Control Valve Opening The LRNBP and FWCF transients are analyzed using Reference 15.E.12-1 and the CLDLP and FCVO transients are analyzed using Reference 15.E.12-4.
DTS:gc:rf/F08013* 15.E.3-9 6/12/85
.: m z. w-I 1 NEUTRON FLUX 2 PERM FUEL CENTER TEMP 3 AVE SURFF CE HEAT FLUX 150*
4 FEEDWATEF. FLOW
. 5 VESSEL S1 ERM FLOW a 100. rudh ,q\
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TIME (SEC) 1 LEVEL (INC H-REF-SEP-SKIRT 2 H R SENSE D LEVEL (INCHES) 3 N R SENSE D LEVEL (INCHES) 200* 4 CORE INLE T FLOW (PCT) -
5 DRIVE FLC 64 1 (PCT) 100 1m y-
- - --w E % g
- 0. ,
~
-100.y*a ' ' ' ' ' ' ' '2. 4. 6. 8.
TIME ISEC)
- GENERATOR LOAD REJECTION W/0 BYPASS FIG 1
CLEVELAND ELECTRIC 104.2% Power 105% Core Flow 15E3-1 t --- r ,--, e- . , - - -- - , . , - , - - - - - , - - - - , , , , - - , - - - - - - - . , , - - - , - , - - , - - - - , , - - - - -
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l 1 VESSEL PF ES RISE (PSI) l 2 STM LINE PRES RISE (PSI) '
3 SAFETY VF LVE FLOW (.)
4 RELIEF Vf LVE FLOW (.)
300. l 5 BYPASS VF LVE FLOW ( ) !
6 TURB STEf M FLOH (PCT) ,
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A 100.
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TIME (SEC)
CLEVELANDEi.ECTRIC
, 104.2% Power 105% Core Flow h.,
1
,-_--,-.-...---e .,c ,,,a -..~,,-y---,-- ,_,--,_..__ ,.---__.,..--_. , .... ,,-...-- ,y,, . .__w.-, , , - - - _ - , _ _ y,.-7 _ __,, -_,,
1 NEUTRON F' LUX 2 PEAK FUEL CENTER TEMP !
A 3 AVE SURFF CE HEAT FLUX 150*
ft 4 FEEDHATEF FLOW 5 VESSEL S1 EAM FLOW 8 100. 'Ti ^ A v
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q' TIME (SEC) 1 LEVEL (IN 4-REF-SEP-SKIRT 2 H R SENS D LEVEL (INCHES) 3 N R SENS D LEVELllNCHES) 200. 4 CORE INLET FLOW 7 CT)
4e A 4 Y ~
M?9 _s
- 0. ,
100.'''''2.
D.
y.
TIME (SEC)
- 6. 8.
FIG GENERATOR LOAD REJECTION W/0 BYPASS 2
CLEVELAND ' ELECTRIC 104.2% Power 73.6% Flow
-.-i.-
1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) 3 SAFETY Vf LVE FLOW te) 300. y REllEF Vf LVE FLOH te) 5 BTPASS VF LVE FLOW te) 6 TURS STEF M FLOW IPCT) 200.
.w %
4 4 -
100.
O. 3@. . . 635 83 h 6 O. 2. 4. 6. S.
TIME ISEC) j N 1 VOID RERdTIVITT 2 DOPPLER FEACTIVITY 3 SCRAM RECTIVITT
- 1. 4 T0iRL REFCTlilTT
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0 la - .
x x 1 A t
- 1. 2. s. u.
TIME (SEC)
FIG GENERATOR LOAD REJECTION W/0 BYPASS CLEVELdNDELECTRIC 15E3 2 104.2% Power 73.6% Flow Cont.
1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 3 AVE SURFF CE HEAT FLUX 150. 4 FEEDWATEF FLOW 4 4 5 VESSEL S1 ERM FLOW
,,p -12e r 100.
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b 3 5 50. b. A_
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- 15. 20.
D. 5. 10.
TIME ISEC) 1 LEVEL (INC H-REF-SEP-SKIRT 2 H R SENSED LEVEL (INCHES) 3 N R SENSED LEVELIINCHES) 150. 4 CORE INLE I FLOW (ftT) 5 5 DRIVE FLC A 1 (PCT) 4 u A 100. E r N 50* W h r
2 '
O.'....I....
- 0, 5. 10. 15. 20.
TIME ISEC) -
FEEDWATER CONTROLLER FAILURE FIG.
CLEVELAND ELECTRIC 15 E 3-3
, 104.2% Power 105% Flow
..nu _ w
~
3' 1 VESSEL PF ES RISE (PSI) g a2 2 STM LINE PRES RISE (PSI) 125. "
, B 3 TURBINE F RES RISE (PSil LA CORE INLE T SUB (BTU /LB) g s 5 RELIEF VF LVE FLOW (PCT) 6 TURB STEF M FLOW (PCT) 2
- 75. <
u u 7 25.
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-25. '. ' -
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TIME ISEC) 1 VOID REn 2 D(1P[i,:Ltf. '
ACT 1VITT 3 5f 62..i F,L U;T lvlT Y
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TIME ISEC) ,
- CLEVELANDILECTRIC FEEDWATER CONTROLLER FAILURE
}{33
. 104.2% Power 105% Flow Cont.
i l
1 NEUTRON FLUX l 2 PEAK FUEL CENTER TEMP 3 AVE SURFF CE HEAT FLUX 150. y FEEDHRTEF FLOW -
9 4 5 VESSEL 51 ERM FLOW 100.
M b E b 8 ; _
b 50. d - -
M : L% 5 E : g I a
' 1 u u
- 0. ' >> '
O. 5. 10. 15. 20.
TIME (SEC) 1 LEVELIINC H-REF-SEP-SKIRT 2 W R SENSFD LEVEL (INCHES) 3 N R SENS!D LEVEL (INCHES) 150.
4 CORE INLE I FLOW (PCT) -
45 45 50.
g 3 t
! -l - <
O.
- 0. 5. 10. 15. 20.
TIME (SEC)
FEEDWATER CONTROLLER FAILURE FIG CLEVELAND' ELECTRIC 15E 3 4 104.2% Power 73.6% Flow
1 VESSEL PF ES RISE (PSI) 125*
3 32
(' /
fA \ 2 STM LINE PRES RISE (PSI) 3 TURBINE F RES RISE (PSI) 4 CORE INLE I SUB (BTU /LB) g s f S RELIEF VF LVE F10H (PCT) 6 TURB STEF M FLOW (PCT) 9
- 75 -
f N f
f4 '
u
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25 o: , ( bhN N '
~
-25. 15, 20.
- 0. 5. 10.
TIME (SEC) 1 VOID REAe d gT A 2 DOPP.4.f.R4EACTIVITY g*
./ \ 3 SCP7. REE CTIVITY
- 11Vlli
{ pTALREF 1
0- #-
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h ~* *: I C :
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- i-2.D. 5. 10. 15. 20.
TIME (SEC)
FIG FEEDWATERCONTROLLERIAILURE CLEVELAND ELECTRIC 5
, 104.2% Power 73.6% Flow ggnt.
15.E.4 Fuel Integrity - Stability The General Electric Company has established stability criteria to demonstrate compliance to requirements set forth in 10CFR50 Appendix A, General Design Criteria (GDC) 10 & 12. These stability compliance criteria consider potential limit cycle response within the limits of safety system or operator intervention and assure that for GE BWR fuel designs this operating mode does not result in specified acceptable fuel design limits being exceeded. Furthermore, the onset of power oscillations for which corrective actions are necessary is reliably and readily detected and suppressed by operator actions and/or automatic system functions. The stability compliance of those GE BWR fuel designs contained in the General Electric Standard Application for Reactor Fuel (GESTAR, Reference 15.E.12-6) is demonstrated on a generic basis in Reference 15.E.12-5 (for operation in the normal as well as the extended operating domain). The NRC has reviewed and approved this in Reference 15.E.12-8 therefore a specific analysis for each cycle is not required.
For operation in the Maximum Extended Operating Domain (MEOD) the stability margin (defined by the core decay ratio) is reduced as power increases for a 1
given core flow. However, the normal realistic operating region has the lowest stability margin at the rated pump speed / minimum valve position flow (minimum forced circulation) which corresponds to about 43% core flow for PNPP (illustrated in Figure 15.E.2-1). Operating at this core flow relative to natural circulation
! results in adequate stability margin f'or the maximum extended operating domain as demonstrated by tests at operating BWRs. Inadvertent operation below minimum forced circulation flows can only occur during transients, e.g., two recirculation pump trip. Operation in the high power / low core flow corner of l
the power flow map is addressed in a set of GE operating recommendations (Reference 15.E.12-7) which have been approved by NRC and will be utilized at l PNPP.
Stability tests were performed in the ME00 region in September and October of 1984 at an overseas BWR6 plant during the initial cycle startup testing. The test objectives were to obtain stability data at high power / flow ratios, to obtain data at reduced feedwater temperature conditions and to evaluate the DTS:gc: rm/F08013* 15.E.4-1 5/13/85 l
r l
l l core behavior when the plant was operated beyond the inception point of limit l cycle oscillations. The tests were conducted under a range of power / flow conditions and data were recorded during the approach to limit cycles, during l limit cycles and during core flow changes once limit cycles were achieved.
l The oscillations observed during the test conditions were compared to those analyzed in Reference 15.E.12-5 and are shown to be bounded by the analyses.
Therefore, consistent with the analyses of Reference 15.E.12-5, large margin to thermal / mechanical limits existed during the test conditions.
)
l l
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i l
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9 i
l DTS:gc:rm/F08013* 15.E.4-2 l 5/13/85 r
- _ _ _ _ _ -M
\
15.E.5 Loss of Coolant Accident Analyses i
A bounding BWR6 Loss of Coolant Accident (LOCA) analysis was performed in the Maximum Extended Operating Domain (ME00) boundary defined ir. Figure 15.E.2-1.
The results are reviewed for applicability to PNPP. It is found that the LOCA analysis results presented in Chapter 6 are adequate to cover the entire ME00 as defined in Figure 15.E.2-1. Therefore, the results meet the 10CFR50.46 limits.
l 1
w i
d DTS: gc: rm/F08013* 15.E.5-1 '
5/13/85
i
, 15.E.6 Containment Response Analysis A containment response analysis is performed with the most limiting postulated recirculation piping line break at the most limiting condition of MEOD for PNPP. The results show that the peak drywell pressure and the peak containment pressure and temperature are still lower than that reported in Chapter 6.
j i
i l
i 1
i i
. i OTS:ge:rm/F08013a 15.E.6-1 5/13/85
15.E.7 Load Impact on Internals 15.E.7.1 Acoustic and Flow Induced Loads The acoustic loads are loads on the vessel internals from propagation of the decompression wave created by a sudden vessel nozzle break. The acoustic loading on the vessel internals is proportional to the total pressure wave amplitude in the vessel upon the postulated break. The additional subcooling in the downcomer resulting from operating in the increased core flow region of MEOD leads to an increase in critical flow and, therefore, in flow induced loads. However, the maximum subcooling in the MEOD is less than the partial feedwater heating operation described in Section 15.0.6. Therefore, it is concluded, based on analysis presented in Section 15.D.6, the reactor internal components have enough margin to handle the acoustic and flow induced loads.
15.E.7.2 Reactor Internal Pressure Difference Loads A reactor internals pressure difference analysis is performed for the increased core flow region of ME00. The increased reactor internal pressure differences across the reactor internals are generated for the maximum core flow at normal, upset, emergency and faulted conditions for the reactor internal impact evaluation.
- 15.E.7.3 Impact on Reactor Internals The reactor internals most affected by pressure under increased core flow conditions are the core plate, guide tube, shroud head, upper shroud, lower shroud, shroud support ring, shroud top guide, fuel channel wall, grid, steam dryer and jet pump. These components are evaluated under normal, upset emergency and faulted conditions. It is concluded that the pressure differences for these and other components during increased core flow operation produce stresses that are within the allowablo limits.
i 1
DTS:gc:rm/F08013* 15 E.7-1
! 5/13/85 l
f l
15.E.8 Flow Induced Vibrations l
l To ensure that the flow-induced vibration response of the reactor internals is acceptable, Perry must undergo an extensive vibration test during initial plant startup in accordance with Regulatory Guide 1.20. PNPP startup test procedure will include flow induced vibration tests at increased core flow region to verify safe ope
- ration as defined in this appendix.
DTS:ge: rm/F08013* 15.E.8-1 5/13/85
15.E.9 Impact on Anticipated Transient Without Scram (ATWS)
An ATWS performance impact evaluation was performed for PNPP in the MEOD.
I Results show that all ATWS consequences from initial ME00 conditions satisfy the criteria for ATWS in Appendix 15C. Furthermore, analysis was also conducted j to assure that there is no unacceptable stability consequences if an ATWS were
- postulated from initial MEOD operating conditions. Therefore, it is concluded i
that ME00 operation is acceptable from ATWS requirements including ATWS stability
{ considerations. '
J I
I L
4 l
1
, DTS:ge:rm/F08013* 15.E.9-1 5/13/85
l I
l 15.E.10 Fuel Mechanical Performance Evaluations were performed to determine the acceptability of PNPP ME00 opera-tion on GE fuel rod and assembly thermal / mechanical performance. Component i pressure differentials and fuel rod overpower values were determined for anticipated operational occurrences initiated from MEOD conditions. These values were found to be bounded by those applied as the fuel rod and assembly design bases and therefore, PNPP MEOD operation is acceptable and consistent with fuel design bases. !
An evaluation was also performed which concluded that fuel channel bypass flow, creep and control blade interference are not impacted by operation in the MEOD.
f DTS:gc: rm/F08013* 15.E.10-1 L 5/13/85 l
15.E.11 Partial Feedwater Heatina (PFH) Operation in the Maximum Extended Operatina Domain This section presents the results of the safety evaluation for operation of PMPP with partial feedwater heating at steady state condition during the operating cycle in the entire MEOD region and beyond the end of cycle in the Increased Core Flow Region (ICFR) of the ME00 as illustrated in Figure 15.E.2-1.
The evaluation is performed for the GE6 fueled PNPP on a equilibrium cycle basis and is applicable to initial and reload cycles operation. The condition of operation are those of continued 100% thermal power operation during and beyond the operating cycle with rated feedwater temperature ranging from 420*F to 320*F in the entire MEOD region and beyond the operating cycle with rated feedwater temperature ranging from 320*F to 250*F in the increased core flow region up to 105% core flow.
All the impact evaluation described in Appendix 150 were reevaluated or reviewed i l
in the entire ME00 for PFH operation. Most conclusions made in Appendix 150 are directly applicable to the PFH operation in the extended region except the abnormal operating transients for which additional increase in operating limit MCPR values is required for PFH operation in the extended regions. Table 15.E.2-1 summarizes the required operating limit MCPR values for PFH modes of operation in various operating region.
15.E.11.1 Abnormal Operatina Transients Two limiting abnormal operating transients (Load Rejection With Bypass Failure and Feedwater Controller Failure) were reevaluated for PFH operation in the extended operating region. The other two limiting transients (100*F Loss of Feedwater Heating and Rod Withdrawal Error) have been discussed in Section 15.E.3 and in 15.0.2 of Appendix D to show that PFH operation in.the extended region is bounded by the results presented in those sections.
t i
DTS:gc: rm/F08013* h 15.E.11-1 5/13/85 L
The Load Rejection with Bypass Failure and Feedwater Controller Failure events were reanaly' zed at: 250*F rated feedwater temperature beyond the end of cycle with 105% rated core flow, 370*F and 320*F rated feedwater temperatures at the end of cycle and also 2000 MWD /T exposure before end of cycle in the ME00 condition. Consistent assumptions and initial conditions in 15.D and in this Appendix are used in the analyses. The transient peak value results are summarized in Table 15.E.11-1 to Table 15.E.11-3. The Critical Power Ratio f
(CPR) results are summarized in Table 15.E.11-4 to 15.E.11-6. The transient responses for the most limiting end of equilibrium cycle cases are presented in Table 15.E.11-6.
The results of the evaluations show that the ACPRs for both the Load Rejection With Bypass Failure event and the Feedwater Controller Failure event exceed the standard operating limit basis. Therefore, a rated operating limit MCPR value of 1.19 (1.20 for reload cores) is required for PFH operation during the operating cycle and beyond the end of cycle in the MEOD for rated feedwater temperature in the range of 420 F and 320*F. However, a rated operating limit MCPR value of 1.21 (1.22 for reload cores) is required for PFH operation beyond the end of cycle in the increased core flow region (core flow > 100% rated) for rated feedwater temperature in the range of 320 F to 250*F.
DTS:gc:rm/F08013* 15.E.11-2 5/13/85
Table 15.E.11-1 Summary of Transient Peak Values Results - Partial Feedwater Heating in MEOD Beyond End of Equilibrium Cycle (a)
Peak Peak Peak Peak Neutron ( ) Dome Vessel Steamline Fdwtr.
Core Flow Flux Pressure Pressure Pressure Temp.
Transient (% NBR) (% NBR) (psig) (psig) psig (*F)
Load Rejection 105.0 286 1192 1220 1194 252 with Bypass Failure Feedwater 105.0 191 1128 1151 1127 252 Controller Failure, Max.
Demand (a) Initial power and heat flux is 104.2% NBR for analysis. Rated feedwater temperature is 250 F.
l 15.E.11-3 DTS:pc/LO40411*
4/19/85 -
i Table 15.E.11-2 Summary of Transient Peak Values Results - Partial Feedwater Heating in ME00 Beyond End of Equilibrium CycleI ")
Peak Peak Peak Peak .
Neutron Dome Vessel Steamline Fdwtr.
Core Flow Flux Pressure Pressure Pressure Temp.
Transient (% NBR) (% NBR) (psig) (psia) psig (*F)
Load Rejection 108.7(b) 304 1203 1233 1211 373 with Bypass Failure 100.0 245 1202 1230 1209 373
{4.8 172 1205 1225 1212 373 Feedwater 108.7(b) 163 1159 1187 1159 373 Controller Failure, Max.
Demand
, 100.0 144 1160 1187 1158 373
- l 74.8 116 1160 1179 1160 373 l Load Rejection 110.0(b) 303 1199 1227 1201 322 with Bypass Failure 100.0 246 1198 1224 1201 322 73.7 162 1200 1218 1201 322 ,
15.E.11-4 DTS:pc/LO40411*
4/19/85
Table 15.E.11-2 Summary of Transient Peak Values Results - Partial Feedwater Heating in MEOD (Continued)
Beyond End of Equilibrium Cycle (a)
Peak Peak Peak Peak Neutron Dome Vessel Steamline Fdwtr.
Core Flow Flux Pressure Pressure Pressure Temp.
Transient (% NBR) (% NBR)(a) ( sig) ( sig) psig (*F)
Feedwater 110.0(b) 219 1151 1177 1151 322 Controller Failure, Max.
Demand 100.0 139 1145 1167 1145 322 73.7 120 . 1149 1167 1148 322 (a) Initial power and heat flux is 104.2% NBR for analysis. Rated feedwater temperature 370 F and 320*F.
(b) Maximum achieveable core flow for the given feedwater temperature.
15.E.11-5 DTS:pc/LO40411*
4/19/85
Table 15.E.11-3 Summary of Transient Peak Value Results - Partial Feedwater Heating in Meod 2000 MWD /T Before End of Equilibrium Cycle (*)
Peak Peak Peak Peak Neutron Dome Vessel Steamline Fdwtr.
Core Flow Flux Pressure Pressure Pressure Temp.
Transient (% NBR) (% NBR) (psig) (psig) psig ( F)
Load Rejection 108.7 104.2 1189 1217 1191 373 with Bypass Failure 74.8 104.3 1192 1212 1196 373 Feedwater 108.7 116.7 1139 1162 1137 373 Controller Failure, Max.
Demand 74.8 115.3 1149 1167 1148 373 Load Rejection 110.0 104.2 1182 1206 1191 322 with Bypass Failure 73.7 104.3 1189 1209 1190 322 Feedwater 110.0 122.6 1117 1141 1116 322 Controller Failure, Max.
Demand 73.7 119.5 1125 1142 1124 327 .
(a) Initial power and heat flux is 104.2% NBR for analysis. Rated feedwater temperature 370 F and 320 F.
15.E.11-6 ~ ~ ' '
DTS:pc/LO40411* i 4/19/85
1
- (
Table 15.E.11-4 Summary of CPR Results - Partial Feedwater Heating in ME00 Beyond End of Equilibrium Cycle (*)
Fdwtr.
Core Flow Temp.
Transient (% NBR) ICPR ACPR MCPR ( F)
Load Rejection 105.0 1.19* 0.13 1.06 252 with Bypass Failure i
Feedwater 105.0 1.21* 0.15 1.06 252 Controller Failure, Max.
Demand 1
- Requires operating limit CPR change.
(a) Initial power and heat flux is 104.2% NBR for analysis. Rated feedwater temperature is 250 F.
I 15.E.11-7 i
DTS:pc/LO40411*
4/19/85
Table 15.E.11-5 ,
Summary of CPR Results - Partial Feedwater Heatup in MEOD Beyond End of Equilibrium Cycle (")
e Fdwtr.
Core Flow Temp.
Transient (% NBR) ICPR(*)(') ACPR MCPR (*F)
Load Rejection 108.7(b) 1.19* 0.13 1.06 373 with Bypass Failure 100 1.18 0.10 1.08 373 74.8 1.27 0.06 1.21 373 Feedwater 108.7( ) 1.18 0.11 1.07 373 Controller Failure, Max.
Demand 100 1.18 0.11 1.07 373 74.8 1.27 0.11 1.16 373 Load Rejection 110.0(b) 1.19* 0.13 1.06 322 with Bypass Failure 110.0 1.18 0.11 1.07 322 73.7 1.27 0.06 1.21 322 15.E.11-8 DTS:pc/LO40411*
4/19/85 .
t i
Table 15.E.11-5 Summary of CPR Results - Partial Feedwater Heating in ME0D (Continued)
Beyond End of Equilibrium Cycle (a)
Fdwtr.
Core Flow Temp.
Transient (% NBR) ICPR(a)(c) CPR MCPR (_F1 Feedwater 110.0(b) 1.19* 0.13 1.06 322 Controller Failure, Max.
Demand 100 1.19* 0.13 1.06 322 73.7 1.27 0.12 1.15 322
- Requires operating limit CPR change.
(a) Initial power and heat flow is 104.2% NBR for analysis. Rated feedwater temperature 370*F and 320*F.
(b) Maximum achievable core flow for the given feedwater temperature.
(c) Based on initial core safety limit of 1.06, for reload cores 0.01 must be added.
15.E.11-9 DTS:pc/LO40411*
4/19/85
- 8 1
1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 3 AVE SURFF CE HEAT FLUX 150. 4 FEEDWATEF. FLOW -
5 VESSEL Sl ERM FLOW 100. Mh E y
S y n , N 1
5 ml - - A
{ 50. ,
5 I'
5 b
- ' ( ' 1 M~ k '
O.
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TIME tSEC) 1 LEVEL (INC H-REF-SEP-SKIRT ,
2 W R SENSED LEVEL (INCHES) 3 N R SENSE.D LEVELfINCHES1 200* 4 CORE INLE T FLOW (PT.T ) ,
5 DRIVE FLC W I (PCTI l 100' MM- ~*
f nm Ay 4
1 31 )
W
- 0. ,
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-10 0.O .' ' ' ' ' ' ' ' ' 2. q. 6. 8.
TIME (SEC)
FIG GENERATOR LOAD REJECTION W/0 BYPASS 15El l-1 CLEVELAND ELECTRIC . 104.2% Power 105% Flow 250*F TFW FFWTR
~
1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) 3 SAFETY VF LVE FLON (e) 300* 4 RELIEF Vf LVE FLOW (+)
5 BTPASS VF LVE FLOW (.)
6 TURB STEF M FLOH (PCT) 200.
1 100.
\ .
D. .3G .... B3 5 8 N 8 3 8.
ab
- 0. 2. 4. 6.
TIME (SEC) 1 VOID PEndTIVITY 2 DOPFLEn I!OCTIVITY 3 SCRAM FC ?l1VITY I* t1 TOInL74r'iivlYY
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TIME (SEC)
FIG GENERATOR LOAD REJECTION W/0 BYPASS CLEVELAND ELECTRIC 15E11 -1 104.2% Power 105% Flow 250 *F FWT Cont.
- x.. c.. . .n . :- -
1 1 NEUTRON F LUX 2 PERK FUEL CENTER TEMP 3 AVE SURFF CE HEAT- FLUX 150' ,',
4 FEEDHATEF FLOW 5 VESSEL S1 ERM FLOW g100.
g _s ,
l.
s b
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\
M w
t
% s
- 4
\ 1 u 1 W u .
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- 10. 15. 20.
TIME (SEC) 1 LEVEL (INC H-REF-SEP-SKIRT 2 H R SENSE D LEVEL (INCHES) 3 N R SENSED LEVEL (INCHES) 150* 4 CORE INLE T FLOW (PCT) 5 DRIVE FLC W 1 (PCT) ku M 100. 3 i
50-r ym' - m 2
5
~.
0 'O. '5. 10.
TIME ISEC)
- 15. 20.
FIG FEEDWATER CONTROLLER FAILURE CLEVELAND ELECTRIC
. 104.2% Power 105% Flow 250*F FWT I
w rw , .-1s,- --._=-.,s,-o-...,--..m- - - - - . , + . - , , - - - , - - - -
1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE IPSI) 12S*
CORE INLE T SUB -(BTU /LB)
RELIEF VF LVE FLOW (PCT) 3 6 RB STEF M FLOW (PCT)
E
_ l 2
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- 0. 5. 10. 15. 20.
TIME ISEC) 1 VOI TIVITY j 24d1PF LER Ii ALTlYITY j* . /\ / 3 SCRAM REf 11VITY 4 TETAL FSTTTiTn
/
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TIME ISEC)
CLEVELAND' ELECTRIC
.104.2% Power 105% Flow 250'F FWT 1 El 1-2 Cont.
-: .26. n.
1 1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 3 AVE SURFf CE HEAT FLUX
~
ISO. 47EEDWATEF FLOW S VESSEL STEAM FLOH o 100. 4A[h 3 - ~
i %
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S DRIVE FLCW 1 (PCT) 2 d-N y
a% 4 5 -
h
- 0. ,
4
-100.
O.' ' ' ' ' ' ' '2' ' . y- 6. 8.
TIME (SEC)
GENERATOR LOAD REJECTION W/0 BYPASS FIG CLEVELAND ELECTRIC
. 104.2% Power 108.7% Flow 370*F FWT
1 VESSEL Pf ES RISE (PSI) 2 STM LINE PRES RISE (PSI) i 3 SAFETT Vf LVE FLOW ( )__
300* 4 RELIEF Vf LVE FLOH ( )
5 BTPASS Vf LVE FLOW (.)
6 TURB STEF M FLOW (PCT) ,
-l IDO. ,,
i 1 ku 100. _
r 7 1
- 0. 3@.. . 63 5 63 'M I3 5 -.
O. 2. 4. 6. 8.
TIME (SEC) 1 l
l 1
1 VOID REAC TIVITY 3*
A V 2 DOPPLER F EACTIVITY 3 SCRAM REF CTIVITT 4 TOTAL REF CTIVIIT ,
l
/
0* I N 9/
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_2, ....i....
O. 1. 2. 3. 4. l TIME (SEC) l l
FIG GENERATOR LOAD REJECTION W/0 BYPASS . l CLEVELAND ELECTRIC 104.2% Power 108.7% Flow 370*F FWT l Co l
K 1
1 NEUTRON F LUX 2 PERK FUEL CENTER TEMP 3 RVE SURFF CE HERT FLUX 150* 4 FEEDWRIEF, FLOW 5 VESSEL SlERM FLOW g 100.
4 h a y 3 b
s _ \
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i
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- 0. e-TIME (SEC) 1 LEVEL (IN K-REF-SEP-SKIRT 2 H R SENS D LEVEL (INCHES) 3 N R SENS D LEVELilNCHES) 200* 4 CORE INLET FLOW (PCT) 5 DRIVE FLC W 1 (PCT) 100. d_
A4 NN 4 -
/ - 3
- 0. .
l
-100.y*' ' ' ' ' ' ' ' '2. 4, 6. B.
TIME (SEC)
FIG.
GENERATOR LOAD REJECTION W/0 BYPASS CLEVELAND ELECTRIC , 104.2% Power 110% Flow 320'F FWT 15 Ell-4
~
. l VESSEL PF ES RTSE (PSI) 2 STM LINE PRES RISE (PSI) 3 SAFETY VF LVE FLOW (e) 300*
4 RELIEF VF LVE FLOW (el 5 BYPASS VF LVE FLOW (*)
6 TURB STEF M FLOW (PCT) 200.
100. Yu '
u D. .36 ,,, 6 35 6'li Ek5 D. 2. 4. 6. 8.
TIME (SEC) 1 V0lD REAdTIVITY
- 1. A Ny 2 DOPPLER F EACTIVITY 3 SCRAM REE CTIVITY 4 TOTAL REF CTlvlTY 1 p -) ./
- -a O-I -
' x\ u !
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-2. -
4.
- 0. 1. 2. 3. ,
TIME (SEC) 1 l
l GENERATOR LOAD REJECTION W/0 BYPASS y)-)_4 CLEVELAND ELECTRIC - 104.2% Power 110% Flow 320*F FWT Cont.
? 1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 1 3 AVE SURFF CE HEAT FLUX 150. y 4 FEEDWATEF FLOW y ,
5 VESSEL Sl ERM FLOW 5 100.
W -
e T" 3 E s 1
j
$ SU' y : .
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- 0. ,,,,i,,,,
- 0. 5. 10. 15. 20.
TIME (SEC) l 1 LEVEL (INC H-REF-SEP-SKIRT 2 H R SENSED LEVEL (INCHES) 3 N R SENSE D LEVELilNCMES) 150. y CORE INLE T FLOW (PCT) 5 DRIVE FLC W1 (PCT) 4 100. 6 N
N ' /
t 50 parmy N .
D. '
- 0. 5. 10. 15, 20.
TIME (SEC)
FIG.
FEEDWATER CONTROLLER FAILURE CLEVELA D ELECTRIC 15 Ell-5 104.2% Power 108% Flow 370*F FWT
n,...~.
F
- l
//
5 RELIEF Vf LVE FLOW (PCT) g h 3 f TURB STEF M FLOW (PCT)
- 75. .
25 th e hs i s
\S \
s\ 4
-25. -
- 0. 5. 10. 15. 20.
TIME (SEC) 1 V01 'TlvlTT A 2 ER F EACTIVITT CTIVITY
/I , CRAM REF I* y y TOTAL REF CTIVITT
' N l
O-1 7" ' / -
nj E 4
- -1.
E :
C : 3 y :
4 *
-2. ~' l - l
- 0. 5. 10. 15. 20.
TIME (SEC)
FEEDWATER CONTR0RER FAILURE 1 Ell-E CLEVELA'ND ELECTRIC 104.2% Power 108% Flow 370'F FWT Cont.
y . . . . w O e e I l 1 NEUTRON F LUX l j
2 PEAK FUEL CENTER TEMP 3 AVE SURFF CE HEAT FLUX 150* 4 4 FEEDWRTEF, FLOW 5 VESSEL S1ERM FLO.W p
5 100.
e E
8 N
es ,
- r 5
e 50
" W ,s E _
l b $
l ....i..., \ 1 u 1 u
- 0. 20.
O. 5. 10. 15.
TIME (SEC) 1 LEVEL (INC H-REF-SEP-SKIRT 2 H R SENS50 LEVELilNCHES) {
3NRSENS$DLEVEL(INCHES) 150* 4 CORE INLET FLOW (PCT) 5 ORIVE FLC A 1 (PCT) 4 9
100.
50- v xmx [ ww D y
- % s D.~'- 15. 20.
- 0. 5. 10.
TIME (SEC)
FIG.
CLEVELANb ELECTRIC FEE 0 WATER CONTROLLER FAILURE ~
. 104.2% Power 110% Flow 320*F FWT
9 f 1 VESSEL PF ES filSE (PSI) 2 STM LINE PRES RISE (PSI) 9 3 TURBINE F RES RISE (PSI) 125* ^ 4 CORE INLE T SUB (BTU /LB) l RELIEF VFLVE FLOW (PCT) J 3 . 6 RB STEFM FLOW (PCT)
B )
~
l 75'
{
N i 1
25 G" 5 5 '
N
-2 5 . 15. 20.
0.~ ' ' ' ' ' ' ' 5 . 10.
TIME (SEC) 1 TIVITY A PLER F EACTIVITY N/\ / SCRAM REF CTIVITY g* y TOTAL REF CTIVITT .
i f
- 0. '*'~ y 5 -1.
s : 4 C :
E b 3 .
4
-2.
' a
- 0. 5. 10. 15. 20.
TIME (SEC) ,
FEEDWATER CONTROLLER FAILURE 1 El'l-6 CLEVELANbELECTRIC Cont.
104.2% Power 110% Flow 320*F FWT
,* )
1 NEUTRON F LUX 2 PERK FUEL CENTER TEMP d 3 RVE SURFF CE HEAT FLUX 150. 4 FEEDWATEF FLOW Ili S VESSEL S1ERM FLOW wm 5, 1 100.
m t e f % l
{5 50. ,
i V b
. 3 1
~
0.
' J ' 1 4 1 -*
0* 2* 4. 6. B.
TIME (SEC) 1 LEVELIINdH-REF-SEP-SKIRT 2 W R SENSED LEVEL (INCHES) 3 N R SENSED LEVELilNCHES) 200. 4 CORE INLBT FLON (PCT)
S DRIVE FLCW 1 (PCT) 100.
- --._JL c 4-L
- 0. .
e
- 10 0. y- 8.
D.' ' ' ' ' ' ' ' ' ' 2. 6.
TIME (SEC)
GENERATOR LOAD REJECTION W/0 BYPASS
~
CLEVELAND ELECTRIC 104.2% Power 74.8% Flow 370*F FWT
>.y. m_.
i, 1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) 3 SAFETT VF LVE FLOW (e) 300* ,
4 RELIEF Vf LVE FLOW (e)
S BYPASS VF LVE FLOW (e) 6 TURB STEF M FLOW (PCT) 200.
< ?
4 100. " 'u~
F 7 U
- 0. 13 tE . . . 63 5 6 E 6[ 5 D. 2. 4. S. 8.
TIME (SEC)
/ 1 VOID READ TIVITT s 2 DOPPLER F EACTIVITY 3 SCRAM REF:TIVITY
- 1. 4 TOTAL REFCTIVITT V
/
1 '
1 - - - .
D. m y S
5 - 1.
E : '
C : .
y y :
- 2. '- \
- 0. 1. 2. 3. 4.
TIME (SEC)
G GENERATOR LOAD REJECTION W/0 BYPASS CLEVELAND ELECTRIC _
. 104.2% Power 74.8% Flow 370*F FWT Cont. !
\
~ - - - - - , - - - . _ - . . _ , , , - , - , - -
_r - - - - - . - - - - - - - - - - - , - - - - - - - -
1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 3 RVE SURFF CE HEAT FLUX ISO
- 4 FEEDWATEF FLOH
[I S VESSEL S1ERM FLOW e 3-% A 9
g W,
4 % s %
z 50*
3 3
~
N ~
d .
h (
b !
- a b
~
"' ' i h 1 -.
- 0. 6. 8.
0* 2* 4.
TIME ISEC) 1 LEVEL (INdH-REF-SEP-SKlRT 2 H R SENSFD LEVEL (INCHES) 3NRSENS!OLEVEll1NCHES) 200. 4 CORE INm61 FLOW (PCT)
S DRIVE FLCW1 IPCT) 100.
Je - ..
u-0.
(
.e
-100. '''' y. 6. 8.
0* 2*
TIME (SEC)
- GENERATOR LOAD REJECTION W/0 BYPASS FIG.
CLEVELdNDELECTRIC 1EE11-8
. 104.2% Power 73.7% Flow 320'F FWT
-_w
! VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) 300* 3 SAFETY VF LVE FLOW to) 4 RELIEF VfLVE FLOW (e)
S BTPASS VF LVE FLOW-le) 6 TURB STEF M FLOH (PCT)
)
200.
100. Nu l O. .3 @ , . . B35 am e gg[ .,
D. 2. 4. 6, 8.
TIME (SEC)
/ 1 1 VOID REAdTIVITY J 2 DOPPLER FEACTIVITY 3 SCRAM REf :TIVITY l 3*
4 TOTAL hEfCiltlT1 ,
V 1
1
- 0. -
x D' G. 1 U .I.
s : -
p :
e -
g :
2, ....i.,,,! 4
- 0. 1. 2. 3. 4.
TIME (SEC)
I 1
FIG. j GENERATOR LOAD REJECTION W/0 BYPASS 15 Ell-8 CLEVELAND ELECTRIC 104.2% Power 73.7% Flaw 320* F FWT Cont. j
1 NEUTRON F LUX 2 PERK FUEL CENTER TEMP 3 RVE SURFF CE HEAT FLUX 150. y y FEEDWATEF FLOW -
.' y 5 VESSEL S1ERM FLOW 1
5 100. L
', W E :
b 3 3 .
E .
I k E :
]V] LN s 4
,,,, 1 u 1 u
- 0. 4
- 0. 5. 10. 15. 20.
TIME (SEC) 1 LEVEllINCH-REF-SEP-SKlRT 2 W R SENSE D LEVEL (INCHES) 3 N R SENSED LEVEll1NCHES) 150*
4 CORE INLE T FLOW (PCT) 5 DRIVE FLC W1 (PCT) 100.
4 50 T-aNe 9 5
- 0. '. ' - *
- 0. 5. 10. 15. 20.
TIME (SEC)
CLEVELAND ELECTRIC FEEDWATER CONTROLLER FAILURE (({j)_g
. 104.2% Power 74.8% Flow 370*F FWT
o f I VESSEL PF ES RISE IPSI) 125* ./
3 2
2 STM LINE PRES RISE (PSI) 3 TURBINE F RES RISE (PSI) 4 CORE INLE T SUB (BTU /LB)
I 5 RELIEF VF LVE FLOW (PCT) g 6 TURB STEF M FLOW (PCT) ya T
7S.
1 4_ )
2S u 3
, N ,
- , f '
t
-h 55 ) B 5 6
-25.
~
c *
- 0. 5. 10. 15. 20.
TIME (SEC) 1 V01 .HjlVITY h 2 rPLER F EACTIVITY
/\ , SCRAM REF CTIVITY
- 1. P { 4 TOTAL REF CTlv1TT i
D. is - -
l V 2
U -1. t
, I :
4 l
C :
W -
y
~
,...i.... d 2*
i
- 0. 5. 10. 15. '
20.
TIME (SEC)
. FEEDWATER CONTf0LLER FAILURE FIG.
CLEVELAND ELECTRIC 15E11-9
. 104.2% Power 74.8% Flow 370"F FWT Cont.
_.-,g_
1 NEUTRON F LUX .
2 PEAK FUEL CENTER TEMP 150* 9 3 RVE SURFF CE HERT FLUX 4 4 FEEDWTTEF FLOW 5 VESSEL Sl ERM FLOW 5 100. '*
5 f W
E E
E w 50. z
$ d
[
3_ _
~
9: -
'a
- 0. ,,,,t....
\
\ 1 u 1 N u .
O. 5. 10. 15. 20.
TIME (SEC) 1 LEVELilNC H-REF-SEP-SKIRT 2 W R SENSED LEVEL (INCHES) 3 N R SENSE D LEVEL (INCHES) 150. 4 CORE INLE T FLOW (PCT) 5 DRIVE FLC W 1 (PCT) 100.
un k
\" ' '
50
, 4 y
~ '
D.''''. 0. 5 10.
TIME (SEC)
- 15. 20.
ON RO N FAILH E CLEVELA'ND ELECTRIC 15 i-if 104.2% Power 73.7% Flow 320*F FWT
.,.r---y - - , - , - . - . , _ , _ . . _ . , . . . - . , . , . _ _ _ . . - - . -
1 VESSEL PF ES RISE IPSI) 2 STM LINE PRES RISE (PSI) 125*
3 TURBINE F RES RISE (PSI) 3 4 CORE INLE I SUB (BTU /LB) 5 RELIEF VF LVE FLOW (PCT)
TURB STEF M FLOW (PCT)
{
- 75. N -
) (
\
- 25. i A
\
- j
- m = ms s e \
l
-25.
0.'''. 5. 10.
TIME ISEC)
- 15. 20.
l 1 TIVITT 1 LER F EACTIVITY
/\ / SCRAM REF CTIVITY 3* 4 TOTAL REf CTivlTT rV 0- #'
a '- x' #
r A
~
~7 l G l
-1.
E _
l E :
g .
~
4 5
c
,,,,t,... B 4 2,
- 0. 5. 10. 15. 20.
TIME (SEC)
J l F FEEDWATER CONTROLLER FAILURE gg CLEVELANDELECTRIC
- 104.2% Power 73.7% Flow 320'F FWT Cont.
'W
15.E.11.2 Other Evaluation, PFH Operation in ME0D All other evaluation described in Appendix 15D for PFH are directly applicable to the PFH operation in the extended operating Region. The 100 F Loss of Feedwater Heating results are bounding for ME0D conditions. The rod withdrawal error analysis described in Appendix 15.D is directly applicable to the ME0D because the bounding generic RWE analysis is performed based on the ME00. The stability criteria are met at the PFH condition in the ME0D. The stability discussion in Section 15.D.3 and 15.E.4 are directly applicable to PFH operation in the ME0D. Both Loss of Coolant Accident and Containment Response Analysis are shown to be bounded by the design values of Chapter 6. Acoustic and flow induced loads on the vessel internal are also demonstrated to be within limits.
Feedwater nozzle, sparger fatigue and system piping are independent of core flow rates. Impact on ATWS, annulus pressurization loads and fuel mechanical performance are also shown to be acceptable for PFH operation in the extended operating region.
)
l l
l I
i DTS:gc: rf/F08013* 15.E.11-11 6/12/85
- _ _---_-- - l
i 15.E.12 References 15.E.12-1 " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" NED0-24154 Oct. 1978.
15.E.12-2 "Three Dimensional BWR Core Simulator" NED0-20953-A, January 1977.
15.E.12-3 Letter, J. S. Charnley (GE) to F. J. Miraglia (NRC), " Loss of Feedwater Heating Analysis", July 5, 1983 (MFN-125-83).
15.E.12-4 R. B. Linford " Analytical Methods of Plant Transients Evaluations for the General Electric Boiling Water Reactor" NED0-10802 April 1973.
15.E.12-5 " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria" NEDE-22277-P-1, October 1984.
15.E.12-6 " General Electric Standard Application for Reactor Fuel" NEDE-24011-P-A, January 1982.
15.E.12-7 "BWR Core Thermal Hydraulic Stability", SIL No. 380 Revision 1, February 10, 1984.
l 15.E.12-8 Acceptance for Referencing of Licensing Topical Report NEDE-24011 Revision 6, Amendment 8, " Thermal Hydraulic Stability Amendment to GESTAR II" NRC's C. O. Thomas to H. C. Pfefferlen, April 24, 1985.
(NOTE: Reference 15.E.12-5 above is the principal supporting document for GE's Topical Report HEDE-24011 Amendment 8.)
DTS: gc: rm/F08013* 15.E.12-1 5/13/85 l