ML20207G699
| ML20207G699 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/31/1986 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20207G695 | List: |
| References | |
| TAC-61357, NUDOCS 8607230146 | |
| Download: ML20207G699 (58) | |
Text
-.
1 GGNS FECDWATER hFATER(5', OllT OF SERbfCE ANALYSIS MAR,CH 1986 i
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i Vreparu' for MISSIS $fPPI PMER & LiS(iT COW?AhY s
GRAND GULF i & 2 NUCLEA'.1 STAT 10NS I
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APPENDIX 15.B TABLE OF CONTENTS NUMBER TITLE PAGE 15.B FeedwaterHeater(s)OutofService 15.B-1 15.B.1 Introduction and Summary 15.B-1 15.B.2 MCPR Operating Limit 15.B-4 15.B.2.1 Abnormal Operating Transients 15.B-4 15.B.2.2 Rod Withdrawal Error 15.B-6 i
15.B.2.3 Operating MCPR Limit 15.B-7
- 15. B. 3 Stability Analysis 1$.B-34 15.B.4 Loss-of-Coolant Analysis 15.B-35 15.B.5 Containment Analysis 15.B-36 15.B.6 Acoustic and Flow Induced Loads Impact 15.B-37 on Internals 15.B.7 Feedwater Nozzle Fatigue Usage 15.B-38 15.B.8 Feedwater Sparger Impact Evaluation 15.B-39
- 15. B. 9 Reactor Protection System Setpoint 15.B-41 j
15.B.10 Miscellaneous Impact Evaluation 15.B-42 l
15.B.10.1 Feedwatcr System Piping 15.B-42 l
15.B.10.2 Impact on Anticipated Transients Without 15.B-42 Scram j
15.B.10.3 Annulus Pressurization Loads Impact 15.B-42 15.B.10.4 Fuel Mechanical Performance 15.B-42 15.B.11 References 15.B-43 i
15.B-1 CS:csc:ac/Il0235*-2 11/13/84
LIST OF TABLES NUMBER TITLE PAGE 4
15.B.2-1 Input Parameters and Initial Conditions for Transient at 320 F Rated FFWTR Operation 15.B-8,9,10 15.B.2-2 Input Parameter and Initial Conditions for Transients at 370*F Rated FFWTR Operation 15.B-11,12,13 15.B.2-3 Summary of Transient Peak Values Results FWH05, EOC1 15.B-14 15.B.2-4 Summary of Transient Peak Values Results FWH05, 2000 MWD /T before E0C1 15.B-15 j
~
15.B.2-5 Summary of Critical Power Ratio Results FWH05, E0C1 15.B-16 15.B.2-6 Summary of Critical Power Ratio Results FWH05, 2000 MWD /T before E0C1 15.B-17 15.B.8-1 Summary of Feedwater Sparger Fatigue Analysis Results for FWH05 Operation 15.B-40 i
4 t
i 15.B-il CS:csc:gc/Il0235*-3 11/13/84
9 LIST OF FIGURES J
NUMBER TITLE PAGE 15.B.2-1 LoadRejectionWithBypassFailure104.2%
Power /109.6% flow 370'F feedwater temperature, E001 15.B-18, 19 15.B.2-2 LoadRejectionWithBypassFailure104.2%
Power /110.9% flow 320*F feedwater temperature, EOC1 15.B-20, 21 15.B.2-3 Feedwater Controller Failure 104.2% Power /
109.6% flow 370'F feedwater temperature, EOC1 15.B-22, 23 15.B.2-4 Feedwater Controller Failure 104.2% Power /
110.9% flow 320*F feedwater temperature, EOC1 15.B-24, 25 15.B.2-5 LoadRejectionWithBypassFailure104.2%
Power /109.6% flow 370*F feedwater temperature, E0Cl-2000 MWD /T 15.8-26, 27 15.B.2-6 LoadRejectionWithBypassFailure104.2%
Power /110.9% flow 320'F feedwater temperature, E0Cl-2000 MWD /T 15.B-28, 29 15.B.2-7 Feedwater Controller Failure 104.2% Power /
109.6% flow 370*F feedwater temperature, EOCl-2000 MWD /T 15.B-30, 31 15.B.2-8 Feedwater Controller Failure 104.2% Power /
110.9% flow 320*F feedwater temperature, EOCl-2000 MWD /T 15.B-32, 33 15.B-iii l
CS:csc:gc/Il0235*-4 11/13/84
15.8 Feedwater Heater (s) Out of Service 15.B.1 Introduction and Summary 1
This appendix presents the results of a safety'and impact evaluation for the operationoftheGrandGulfNuclearStations(GGNS)duringtheinitialoperat-ingcyclewithfeedwaterheater(s)outofservice. The evaluation supports operation within the power flow region as illustrated in Figure 4.4-5 of Chapter 4 of GGNS FSAR. This evaluation is performed for the GE-6 fueled GGNS at initial cycle with a target Haling end of. cycle exposure distribution..The condition of operation is that of continued'100% thermal power operation during the normal operation cycle with a maximum rated feedwater temperature reduction j
of100'Fduetofeedwaterheater(s)outofservice. This evaluation justifies GGNS continued operation during the initial operating cycle between 420'F and 320*F rated faadwater temperature at rated power with some required Technical Specification modifications.
I l
Operation at a reduced feedwater temperature occers in the event that certain stage (s)orstring(s)orindividualheatersbecomeinoperable.
Loss of feedwater heating from the highest pressure heaters would result in the highest temperature'
]
reduction.
Loss of heating from the low pressure heaters would result in only i
a slight reduction of feedsater temperature.
Chapter 15 has already evaluated l
the transient response for the worst feedwater temperature loss up to 100'F l
during an operating cycle due to inoperable out of service of unavailable heater stages.
Therefore, this appendix will justify the operation for GGNS f
with the rated feedwater temperatures between 420'F and 320'F due to inoperable feedwaterheater(s).
e l
ThisevaluationistermedFeedwaterHeater(s)outofservice(FWHOS)inthe remaining content of this section. TheonlyadjustmentfortheFWHOSoperating l
condition is to change the RPS scram function on the turbine stop valve closure l
and the turbine control valve closure to assure that the scram bypass is consistent with the 40% of rated power in FWHOS conditions. The technical specification should also be modified to increase the operating limit MCPR by 0.01 for operation between 370'F and 320'F rated feedwater temperature.
No operating limit MCPR change needs to be made for operation between 420'F and 370*F rated feedwater temperature."
CS:csc/110236*
15.B-1 12/7/84
The following evaluations and conclusions resulted:
(a) The abnormal operating transients in Chapter 15 were reevaluated at rated feedwater temperature of 320*F and 370'F to determine the required operat-l ing MCPR limits for FWHOS operation.
Increased core subcooling during FWHOS resulted in the feedwater controller failure become the most limit-ing event which requires a 0.01 increase in operating limit MCPR (OLMCPR) for rated feedwater temperature operation between 370'F and 320'F.
No operating limit MCPR change is required for operation between 420 'F and 370'F rated feedwater temperature.
(b) It is determined that the fuel mechanical limits are met during FWHOS operation under stecdy state and anticipated operational occurrences.
(c) The Loss of Coolant Accident (LOCA) and containment response as described in Chapter 6 were reevaluated for FWHOS operation.
It is found that the normal feedwater temperature analysis adequately bound those with FWHOS conditions.
l (d) Fuel integrity thermal-hydraulic stability was evaluated with respect to GeneralDesignCriterion12(10CFR50,AppendixA).
It is shown that the FWHOS operation satisfies the stability criterion and fuel integrity is not compromised.
(e) The effect of acoustic and flow induced loads on the reactor shroud, shroudsupportandjetpumpswereanalyzedtoshowthatthedesignlimits are not exceeded.
The effect of FWHOS on feedwater nozzle and sparger fatigue usage factor was determined.
It'was found that the increased fatigue usage on the feedwater nozzle adequately meets the acceptance criterion for unlimited operation up to 40 years. The increased fatigue l
usage on the feedwater sparger meets the acceptance criterion with some limitation on the maximum allowable number of days for FWH05 operation.
l These specific FWH05 operation time limits are of economic concern only.
l l
CS:csc/Il0236*
15.8-2 l
12/7/84 l
l
+ - - -, - - - - - _ - - -
(f) The turbine stop valve and the turbine control valve scram bypass setpoints intheReactorProtectionSystemshouldbereadjustedtobeconsistentat 40% rated power with reduced Feedwater temperature conditions.
The recom-mended setpoints are 21.0
- 0.5% of the calibrated span for 420'F to 370'F rated feedwater temperature operation and 18.0
- 0.5% of the calibrated span for 370'F to 320'F rated feedwater temperature operation.
There are also other impact evaluations performed such as feedwater system piping, annulus pressurization load analysis, and Anticipated Transients WithoutScram(ATWS)tojustifyFWHOSoperation. These evaluations concluded that the standard operation design is adequate for FWHOS operation.
I CS:csc/Il0236*
15.B-3 12/7/84
l 4,
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15.B.2 MCPR Operating Limit 15.B.2.1 Abnormal Operating Transients All abnormal operating transients described in Chapter 15 were examined for FWHOS operation. Three limiting abnormal operating transients wer'e reanalyzed in detail with a bounding BWR 6 standard plant at 104.2% power at both 75% and 110% core flow and reevaluated for GGNS at about 110% core flow. These analyzed transients are:
a)
GeneratorLoadRejectionwithBypassFailure(LRNBP) b)
FeedwaterFlowControllerFailure(FWCF) c)
Loss of 100'F Feedwater Heating (LFWH) l The GGNS specific LRNBP and FWCF transients were evaluated at 104.2% power and about 110% core flow with rated feedwater temperature at 320'F and 370 F at end of cycle 1 and 2000 MWD /T exposure before end of cycle.
The 75% core flow analysis was not performed for GGNS because the bounding BWR 6 analysis performed at both 370*F and 320*F rated feedwater temperatures has shown significant margin existed at this condition.
Plant heat balance, core coolant hydraulic and nuclear transient data were developed and used in the analysis.
The initial conditions for the analysis points are presented in Tables 15.B.2-1 and 15.B.2-2.
The end of cycle exposure point with all control rods fully withdrawn is a limiting point in the cycle with the worst scram worth reactivity charac-teristics. The 2000 MWD /T before end of cycle exposure point is chosen as an analyzed point because it is close enough to end of cycle such that the scram characteristics have not been significantly improved relative to earlier points in the cycle and the void reactivity characteristic is different than end of cycle.
Scram characteristics are significantly improved at exposure lower than this point and the transient responses will be bounded by the two points 1
analyzed.
CS:csc/Il0236*
15.B-4 10/23/84
i The computer model described in reference 15.B.11-1 was used to simulate both of these events.
The transient peak value results and critical power ratio results are summarized in Tables 15.B.2-3 to 15.B.2-6.
The transient responses for these end of cycle and mid cycle cases are presented in Figures 15.B.2-1 to 15.B.2-8.
_The bounding BWR 6 analysis and the GGNS specific analysis shown on Tables 15.B.2-5 and 15.B.2-6 indicate that the ACPR for the worst feedwater controller failure transient exceeds the standard operating limit MCPR by 0.01 for opera-tion at reduced rated feedwater temperature down to 320*F.
Therefore, the current. Technical Specification rated MCPR operating limit of 1.18 needs to be modified to the new rated OLMCPR of 1.19 between 370*F and 320'F rated FWHOS operation. The OLMCPR value does not need to be changed between 420 and 370*F rated FWHOS operation.
Note that the 1.19 OLMCPR value is to be conservatively applied to FWHOS operation between 420'F and 320*F in the Maximum Extended Operating Domain (MEOD) based on the bounding BWR 6 evaluation.
[,
Lower initial operating pressure and steam flow rate provide better overpressure prciection for the most limiting MSIV closure (flux scram) event during FWH05 operation. Tables 15.B.2-3 and 15.B.2-4 also indicates that the peak pressures for the LRNBP and FWCF events analyzed are below the ASME code value of 1375 psig.
Hence, it is concluded that the pressure barrier integrity is maintained under l
FWHOS operation conditions.
l The 100'F loss of feedwater heating transient was evaluated for a bounding BWR 6 plant at initial feedwater temperatures of 250*F (to bound all FWHOS operation) and 420'F at 104.2% power, 100% core flow at the end of the cycle using the computer model described in reference 15.B.11-3 and methodology described in reference 15.8.11-4.
Results show that the 100'F LFWH has less effect on colder feedwater than on the normal feedwater temperature.
The ACPR result for the 100'F loss initiated from 250*F is bounded by the 420*F initia-tion case. The generic LFWH analysis described in reference 15.B.11-4 concluded that this event is insensitive to initial core flow.
It is less likely that a 100*F loss would occur at an initial feedwater temperature of 370'F or 320*F during FWHOS operation.
A generic statistical LFWH analysis using the same model and methodology described above utilizes a large data base of LFWH cases CS:csc/Il0236*
15.B-5 3/5/85
.-. J.---
to generate the 95% probability, 95% confidence bounding ACPR value for the normal 420*F condition.
This data base includes transient responses at different exposure points throughout the operating cycle.
Therefore, the loss of feedwater heating analysis for FWHOS is adequately bounded by the 420'F ACPR results.
The pressure controller downscale failure event does not need to be evaluated.
The GGNS specific steam bypass and pressure regulator system has been reviewed to show that no possible faults in the electronics that could result in the pressure regulator failure downscale with a single failure.
The system is made of three independent channels that are compared at various points during single processing so that a single failure is one channel would be detected.
15.B.2.2 Rod Withdrawal Error The rod withdrawal error (RWE) transient documented in chapter 15 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 from both rated and off rated conditions over the entire operating region.
Therefore, this analysis covers a wide variety of feedwater temperatures and core subcooling as different off rated conditions are included in the data base.
The 95% probability 95% confidence values from this statistical data base are used to develop the Rod Withdrawal Limiter (RWL) system setpoints to protect against a rod withdrawal error.
The rod withdrawal error analysis does not need to be evaluated for FWHOS at end of cycle because all control rods will be fully withdrawn.
A RWE analysis was performed at 2000 MWD /T before end of cycle to examine the effect of the initial feedwater temperature.
An initial condition of 250*F was used to bound all FWHOS operation.
Results show that ACPR resulting from the worst 2 feet of withdrawal for the 420'F and 250*F feedwater temperature are identical.
Therefore, the ACPR values initiating from 250'F feedwater temperature condition fall within the statistical data base used to establish the RWL system setpoints.
The change in linear heat generation rate is bounded by the fuel mechanical design bases.
Therefore, it is concluded that operating limit MCPR does not need to be increased due to RWE for FWHOS operation.
CS:csc/Il0236*
15.B-6 3/5/85
15.B.2.3 Operating Limit MCPR For rated FWH05 operation between 370*F and 320*F the rated OLMCPR needs to be increased by 0.01 (to 1.19 for the initial core). The off-rated power dependent MCPR, limits also need to be increased by 0.01 to conservatively cover FWHOS operation at this rated temperature range.
The off rated flow dependent MCPR limits do not need to be changed because the current MCPR limit curve f
f "Was generated based on a steepest power flow rod line to protect against the recirculation flow runout transient.
A power flow rod line was generated for the 320*F rated initial condition.
It shows that the slope of this rod line is bounded by the current design basis rod line.
Therefore, the current MCPR f limits are valid for FWH05 operation.
J CS:csc/Il0235*
15.B-7 3/5/85
Table 15.B.2-1 i
Input Parameters and Initial Conditions for
. Transients and Accidents for FWHOS, 320*F FWT 104.2% Power, 110.9% Flow 1.
Thermal Power Level, MWt 3994 (104.2% rated)
Analysis Value 2.
Steam Flow, alb per hr 15.16 Analysis Value 3.
Core Flow, alb per hr 124.8 4.
Feedwater Flow Rate, alb per hr 15.16 Analysis Value 5.
Feedwater Temperature, *F 320 6.
Vessel dome pressure, psig 1010 7.
Core exit pressure, psig 1020 8.
Turbine Bypass Capacity, % NBR 35 9.
Core Coolant Inlet Enthalpy 517 Btu per ;c
- 10. Turbine Inlet Pressure, psig 941 11.
Fuel Lattice 8x8R
- 12. Core Leakage Flow, %
10.65 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 (-)t/*F 0.132(a)
Analysis Data C5:csc/I1O236*
15.B-8 3/5/85
Table 15.B.2 (Continued)
- 16. Void Coefficient (-)t/% 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, %
36.0
- 18. Jet Pump Ratio, M 2.25
- 19. Safety / Relief Valve Capacity, % NBR 91125 psig 100.6 Manufacturer Dikker Quantity Installed 20
?O.
Relief Function Delay, seconds 0.4 21.
Relief Function Response Time Constant, sec.
0.1
- 22. Setpoints 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 Analysin Setpoint (122x1.042), % NBR 127.1 25.
High Pressure Scram Setpoint, psig 1095
- 26. Vessel Level Trips, Feet Above Separator Skirt Bottom Level 8 - (L8), feet 5.88 Level 4 - (L4), feet 4.03 Level 2 - (L3), feet 2.16 Level 2 - (L2), feet
(-) 2.182
- 27. APRM Thermal Trip Setpoint, % NBR 118.8 28.
RPT Delay, seconds 0.14 CS:csc/Il0236*
15.B-9 10/23/84
i Table 15.B.2 (Continued)
- 29. RPT Inertia Time Constant for Analysis, sec.
5
- 30. Total Steamline Volume, ft3 4358 (a) These values for Reference 15.B.11-2 analysis only.
Reference 15.B.11-1 values are calculated within the code.
CS:csc/Il0236*
15.B-10 10/23/84
Table 15.B.2-2 Input Parameters and Initial Conditions for Transients and Accidents for FWHOS, 370*F FWT 104.2% Power,109.6% Flow t
1.
Thermal Power Level, MWt 3994 (104.2% rated)
Analysis Value 2.
Steam Flow, alb per hr 16.10 Analysis Value 3.
Core Flow, alb per hr 123.3,
l 4.
Feedwater Flow Rate, alb per hr 16.10
~~
Analysis Value 5.
Feedwater Temperature
- F 370 6.
Vessel dome pressure, psig 1020 7.
Core exit pressure, psig 1031 8.
Turbine Bypass Capacity, % NBR 35 9.
Core Coolant Inlet Enthalpy 523 Btu per 1b
- 10. Turbine Inlet Pressure, psig 944 11.
Fuel Lattice 8x8R 12.
Core Leakage Flow, %
10.65 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 (-)$/'F 0.132(a)
Analysis Data CS:csc/Il0236*
15.B-11 10/23/84 1
I
.,---r.y v_
Table 15.B.2-2 (Continued)
- 16. Void Coefficient (-)(/% Rated Voids Analysis Data for Power Increase Events 14.0 (a)
Analysis Data for Power Decrease Events 4.0 (a) 17.
Core Averrge Rated Void Fraction, %
38.0
- 18. Jet Pump Ratio, M 2.25
- 19. Safety / Relief Valve Capacity, % NBR 91125 psig 100.6 Manufacturer Dikker Quantity Installed 20 20.
Relief Function Delay, seconds 0.4 21.
Relief Function Response Time Constant, sec.
0.1
- 22. Setpoints 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 127.1 l
25.
High Pressure Scram Setpoint, psig 1095 26.
Vessel Level Trips, Feet Above Separator i
Skirt Bottom l
Level 8 - (L8), feet 5.88 Level 4 - (L4), feet 4.03 Level 3 - (L3), feet 2.16 Level 2 - (L2), feet
(-) 2.182
- 27. APRM Thermal Trip
)
Setpoint, % NBR 118.8 l
28.
RPT Delay, seconds 0.14 l
CS:csc/Il0236*
15.B-12 10/23/84 l
4 Table 15.B.2-2 (Continued) 29.
RPT Inertia Time Constant for Analysis, sec.
5
- 30. Total Steamline Volume, ft3 4358 I
(a) These values for Reference 15.B.11-2 analysis only.
Reference 15.B.11-1 values are calculated within the code.
l CS:csc/Il0236*
15.B-13 10/23/84
Table 15.B.2-3 Summary of GGNS Transient Peak Values Results - FWHOS - EOCl(a)
Peak Peak Peak Peak Neutron Dome Vessel Steamline Fdwtr Core Flow Flux Pressure Pressure Pressure-Temp.
(% NBR)
(% NBR)
(psig)
(psig)
(psig)
( F)
LoadRejection 109.6 (b) 162 1194 1224 1195 370 with Bypass Failure Feedwater 109.6(b) 120.3 1150 1175 1149 370 Controller Failure, Max.
Demand LoadRejection 110.9(b) 166 1188 1218 1189 320 with Bypass Failure Feedwater 110.9 (b) 125 1128 1152 1127 320 Controller Failure, Max.
Demand (a) Initial power is 104.2% NBR for analysis.
(b) Maximum achievable core flow for the given feedwater temperature.
4 CS:csc/Il0237*
15.B-14 10/23/84
Table 15.B.2-4 Summary of Transient Peak Value Results - FWHOS 2000 MWD /T Before EOCl(a)
Peak Peak Peak Peak Neutron Dome Vessel Steamline Fdwtr Core Flow Flux Pressure Pressure Pressure Teb.
(% NBR)
(% NBR)(a)
(psig)
(psid (psig)
(
LoadRejection 109.6 (b) 104.3 1185 1216 1189 370 with Bypass Failure Feedwater 109.6(b) 118.2 1136 1160 1135 370 Controller Failure, Max.
Demand Load Rejection 110.9 (b) 104.3 1177 1205 1184 320 with Bypass Failure Feedwater 110.9(b) 124.3 1112 1137 1111 320 Controller Failure, Max.
Demand Initial power is 104.2% NBR for analysis.
(a) Maximum achievable core flow for the given feedwater temperature.
(b)
CS:csc/110237*
15.B-15 10/23/84
Table 15.B.2-5 j
Summary of CPR Results - FWH05 - EOC1 Fdwtr.
Core Flow Tem Transient
(% NBR)
ICPR(b)
ACPR MCPR (F
LoadRejection 109.6(a) 1.18 0.05 1.13 370 with Bypass Failure Feedwater 109.6(a) 1.18 0.11 1.07 370 Controller Failure, Max. Demand LoadRejection 110.9(a) 1.18 0.05 1.13 320 with Bypass Failure Feedwater 110.9(a) 1.18 0.12 1.06 320 Controller Failure, Max. Demand (a) Maximum achievable core flow for the given feedwater temperature.
(b) Based on initial core safety limit of 1.06, for reload cores 0.01 must be added.
NOTE:
Option A adders included.
CS:csc/Il0237*
15.B-16 10/23/84
Table 15.B.2-6 Summary of CPR Results - FWH05 - 2000 MWD /T Before EOC1 Feedwater Core Flow Temperature Ib)
(% NBR)
( F)
LoadRejection 109.6 (a) 1.18 0.05 1.13 370 with Bypass Failure Feedwater Controller 109.6 (a) 1.18 0.11 1.07 370 Failure, Max.
Demand LoadRejection 110.9 (a) 1.18 0.05 1.13 320 with Bypass Failure Feedwater Controller 110.9 (a) 1.19**
0.13 1.06 320 Failure, Max.
Demand (a) Maximum achievable core flow for the given feedwater temperature.
{b)Basedoninitialcoresafetylimitof1.06,forreloadcores0.01mustbeadded.
- Requires operating limit CPR change.
NOTE: Option A adders included.
i CS:csc/Il0237*
15.B-17 i
10/23/84
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FIGURE :
l POWER & LIGHT
' 320'F TFW EOC1 15.B.2 l l
i l
l 15.B-20 l
e 1 VESSEL PF ES RISE (MI) 2 STM LINE PRES RISE (PSil 1
300*
3 SAFE 11 VI LVE FLOW (el i HELIEF VfLVE FLOW (el t
5 BTPRSS VFLVE FLOW I.)
6 TURB STEf M FLOW IPCT) 200.
i b
W
\\ "-
100.
l i
3[
j g,,f.30....,
4.
b.
U.
835 E
G4 6
4 U.
2.
TIME (SEC1 1
/
I VnID Rinf TIVITY
- ! Dut'11Ill I f0C11VITY g*
1A /
'l ?.iIVIH Id's CllVITY Gulictn.) OTVITT
)
/
0.
I' n
1.
E
(#
W W
?
- 2. w.i....
D.
1, 2.
3.
4.
TIME (SEC)
I I
MISSIS $1PP2 LOAD REJECTION WITH BYPASS FAILURE,104.2% POWER,110.9% FLOW, FIGURE i
POWER 6. LIGHT 320'F TFW EOCl 15.6.2-(Cont.i i
15.B,21
1 NEUTRON f LUX 2 PERh Fil[I CLNTER TEMP 3 AVE SORfl CC HLAT FLUX 150* 4 81 FEDRRID YCOs ~
~
y 5 VESSEL 51 EON FLOW 1
F g
^
E 100.
[
b e-E 50.
b W
a s
~
D.~'-
1 u
1 u
D.
5.
10.
15.
20.
. TIME (SEC) l i
I LEVEL (INC H-REF-SEP-SKIRT 2 H R SENSF 0 LEVFL(INCHES) 150*
3 N R SEN5f 0 1.EVfL(INCHES 1 4 CORD ING TTL' Oil (PCT) 5 DRIVE FLt i4 1 (PCT) u 5
100.
%\\qc
/
50' Iy#%\\[Lu j y
E s-D.
4 0.
5.
10.
15.
20..
TIME (SEC1 MISSISSIPPI FEEDWATER CONTROLLER FAILURE 104.2% POWER,109.6% FLOW, FIGURE POWER & LIGHT 370*F TFW EOC1 15E.2-:
15 B-22
! VESSEL Pr ES RISE (PSI) j 2 STM LINE. PRES RISE (PSI)
/>
3 TURBINE F RI:S 111SE (PSI) 125*
3 3
%:0RGNTl I 50tndTD/LB) 5i LIEF Vi I.VE FLOW IPCT) 6T STEF M FLON IPCT) g j
vs.
N h
u 25.
M 9
F, f
i 3 5 i
5 i
45,0.
~.
S.
10.
15, 20.
TIME (SEC) 1 1 V010J1Enf TTVITY 2 IlillPl. fit I l liri lVITY ll i I L
T T-1
/
A 0*
'y g.
I
-1.
(
k W
4 g
I 4
....t....
.e D.
5.
10.
15.
20.
TIME ISEC) l L
MISSISSIPPI FEEDWATER CONTROLLER FAILURE.104.2% POWER,109.6% FLON, FIGURE MER 8 LIGHT 370*F TFW EOC1 15.B.2 l
(Cont.
t
~
15.B-23
I NEUTRON P LUX 2 PERK FUEL CENTER TEMP 3 RVE SilRFI CE If.RT FLUX 150* M y FEEDWRILI TLOW 5 VESSEL SI CRM FLOW Zl i
_i l
a b
I i
.g '
s X
s s
'U U
^
D.''''.
0.
5 10 15.
20.
TIME ISEC) 1 LEVELiINC H-REF-SEP-SK1RT i
2 H R SENSC D LEVELiINCHES) 150*
3 N R SENM il LEVEL.IINCHES) 11 CORE IRI.I UI 0W7fT1 5 DRIVE FLL H 1 (PC1) i u
5 100.
W%
N X
/
go'If NwM u
~
s m
(
]
0.
0 5
10.
15.
20.
' TIME (SEC)
MISSISSIPPI FEEDWATER CONTROLLER FAILURE,104.2% POWER,110.9% FLOW.
FIGURE POWER & LIGHT 320*F TFW EOC1 15.B.2 t 15.B-24
f
\\ l VESSEL Pits RISE (PSI)
STM LINE 4' TIES RIEE (PSI) f J
T
-DINF F IV.S AlSE (PSII 125.
y gg fA!,i T3tTli"IIITij/LB1 5
- IEF W t.YE FLOW IPCT) 3 6i STU N FLDH (PCT)
- s o
n.
N x
x 25.
I ii 1h 5
B E
E
\\5 3
-25.
8
~
0.
5.
10.
15.
20.
TIME (SEC) 1 vnirtM0f flI'ITY
- ',,lilVTI.lill I ffIC1IVITT I*
/
4 Sr.linM I:l~1 i.livil)
I 1tl'IllC~liCl dlTVI'li
.,f l
/
t l
}
ll D* ' *N l
ff E
b "I'
/
i 4
i if k
l W
Y
{
\\
-2.
0.
5.
10.
15.
20.
TIME ISEC)
MISSISSIPPI FEE 0 WATER CONTROLLER FAILURE.104.2% POWER,110.9% FLOW.
FIGURE POWER & LIGHT 320*F TFW E0Cl 15.2-4 (Cont.)
15.B-25
A l
1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 3 RVE SURFF CE HERT FLUX 150.
4 FEEDWRl G TGW 5 VESSEL S1 ERM FLOW 100 i
b e-5mf b
1 D.
E
'I' I
1
^
O.
2.
4.
6.
8.
TIME ISEC) i I
1 LEVELtINC H-REF-SEP-SKIRT 2 H R SENSE D LEVELilNCHES) 3 N R SENSE D LEVELIINCHES) 200*
4 CORE ]NLf i FLOW (PCT)
NW
_.u u
2, s D.
i E
-100.
I 0.
2.
4.
6.
- 8. ~
TIME ISEC) 9 MISSISSIPPI GENERATOR LOAD REJECTION WITH BYPASS FAILURE,104.2% POWER.
FIGURE POWER & LIGHT 109.6% FLOW 370*F TFW EOCl-2K 164D/T
- 15. B. 2-1 l
14J?L-fd
1 VESSEL PF ES RISE (PSI) 2 STM LINE PRES RISE (PSI) g*
3 SAFETY Vf LVE FLOW (e) te RETTET Vf LVE FLOW (e) 5 DTPASS VF l.YE FLOW l 1 6 TURB STEFM FLOW IPCT) l 200.
l 100*
y p
u 0.Nats....
E3 E E
- u s [5 4
0.
2.
u.
s.
s.
TIME ISEC) 1 VOID Dini TIVITY
' Dt1PI't.f11 I i.nl:11VITY g*
.I !Elliitt liff Cl1V11Y 11 lulHL lill t: l l VT1'T"~
1 D.
1 4
^
f m,*
w>
-1.
E i
C W
E 2, 2........ I D.
1.
2.
3.
4.
TIME ISEC) i MISSISSIPPI GENERATOR LOAD REJECTION WITH BYPASS FAILURE,104.2% POWER.
FIGURE POWER & LIGHT 109.6% FLOW, 370*F TFW EOC1-2K WD/T
- 15. B. 2 (Cont.
15.B-27
i 1 NEUTRON F LUX 2 PEAK FUEL CENTER TEMP 150*
3 RVE SURff CE HERT FLUX 4 FEEDHRTEF FLOW 5 VESSEL S1 ERM FLOW
~
~ 100
- M h
\\
5 50.
~
1 1
-- ^4M O.
I 0.
2.
TIME (SEC) 1 LEVEL (INCH-REF-SEP-SKIRT 1
2 H R SENSE D LEVEL (INCHES) i 200*
3 N R SENSE D LEVEL (INCHES) 4 CORE INLE T FLOW (PCT) 5 DRIVE FLCW I (PCT) i i
100 33 r
u N
1 y,
.s 0.
-100.~
0.
2.
4.
6.
8.
TIME ISEC)
MISSISSIPPI LOAD REJECTION WITH BYPASS FAILURE 104.2% POWER,110.9% FLOW.
FIGURE POWER & LIGHT 320'F TFW EOC1-2K MWD /T 15.B.2-f l
I 15.B-28
1 VESSEL PFES RISE (PSI) 2 STM LINE PRES RISE (PSI) 3 SAFETY VF LVE FLOW t.)
300*
4 RELIEF VF LVE FLOW ( )
S BYPASS VF LVE FLOW t.)
6 TURB STEF M FLOW (PCT) 200.
i l
100' wr u
o, atEr....
63 5 83 Lu B
5 O.
2.
M.,,,,,,,,,,,,6.
8.
sarm. x.u T'
I VOID REAC TIVITY 2 DOPPLER F CACTIVITY 3 SCHAM Rff CTIVITY 3*
i 4 TOTAMI CTIVITY 9
/
1 D.
4 i
3 1
5
-1.
)
t E
W E
i 2.
....n.
I D.
1.
- 2..
3, g.
TIME ISEC1 MISS!$51PPI LOAD REJECTION WITH BYPASS FAILURE,104.2% POWER,110.9% FLOW.
FIGURE ll POWER & LIGli 320'F TFW EOC1-2K WD/T '
15.B.2-(Cont.'
i 15 0-29
-=
1 NEUTRON P LUX 2 PERK FilEl CENTER TEMP 150.
3 RVE SilRF5 CE HEAT FLUX U
4 4 FEEDWATEF FLDW i
~
f 5 VESSEL 51 ERM FLOW i sf Ai 100.
E g-50.
N
)
O.
1 u1 ut
~
0.
5.
10.
15.
20.
TIME ISEC) 1 LEVELIINC H-REF-SEP-SKIRT 2 N R SENSE D LEVEL (INCHES) 150.
3 N R SENSE D LEVEL (INCHES)
~
4 CORE INLl i FLOld (PCT) 5 ORIVE FL[ i4 1 (PCT) u 5
100.
\\
50.
V
%'db 0.
a 0.
5.
10.
15.
20.-
l TIME ISEC)
MISSISSIPPI FEEDWATER CONTROLLER FAILURE.104.2% POWER 109.6% FLOW.
FIGURE POWER & LIGHT 370*F TFW EOC1-2K MWD /T 15.B.2-)
i 15.B-30
. -.ms :.m a
SSEL PF ES RISE (PSI) 2 LINE PRES RISE IPSIt 3T BINE F RCS RISE (PSI) 125*
4 CDR l'NL7 TTutNu/LB) 5 RELI F VF LVE FLOW (PCT) g 6 TURB TES M FLON (PCT) 75.
=
x 23*
N
- iz -
- !s
\\
s e
a
~-25,0.
5.
10.
15.
20.
TIME (SEC1 e
1 VOID Rent TI WV 2 Otir1]?.IH iT. LTIVITT
. eNT>itt r;rt i.TIV1TY A
s t f-inmimi Twat 2-l yJ 1 *A L <
0.
I i
1' 2
(
D
-1.
E
~.
C W
W 2.
N D.
5.
10.
15.
- 20.
TIME ISEC1 MISSISSIPPI FEEDWATER CONTROLLER FAILURE 104.2% POWER,109.6% FLOW.
FIGURE POWER & LIGHT 370'F TFW EOCl-2K PWD/T
- 15. B.2 (cont.
1 NEUTRON P LUX 2 PERK FLIEl CENTER TEMP e
3 RVE Sur#8 CE HEnT FLUX 150-4 FEEDWA16 T G iH u
h 5 VESSEL $1 CRM FLOW L'sa y
100.
b i
1 50-3 s
5.
[
b U'
M'*
D."'-
0.
5.
10.
15.
20.
TIME.ISEC) 1 LEVELIINC H-REF-SEP-SKinT 2 H R SENSI 0 LEVEL (INCHL!il 3 N R SENSI in I.EVFL(INCHES) 150*
i 11 COME IN"LI ITL'OA~TFtT1 5 DRIVE FLt 14 1 IPCT) 100.
i 50' If\\NyMN u
7
.' C E
D.
=
0.
5.
10.
15.
20.-
TIME.15EC)
MISSISSIPPI FEEDWATER CONTROLLER FAILURE.104.2% POWER,110.9% FLOW.
FIGURE POWER & LIGHT 320*F TFW EOC1-2K eld /T 15.B. 2-E
- 15. R-32
-l 1 VESitL Pr ES MISE (PSI) 2 STM i
3Tm t i;Es n15E (PSil r
1 CD IC 1 tiDii"[9TU/LB)
STEH1 FLOW (PCT) s I
9
. "15.
T u
j
\\
4 y
t N
25.
3 3
)
.ow g
g g
NN
\\\\
<=
-25 0.
5.
10.
15.
20.
TIME 1SEC) 1 VOID TT d !.uiT'. "
. g.. r,i in
.;.n ID*1TY n
_ /":.l.it::il f ' el I lY l "i I*
fj it'lilliY ~l
. I t'. l'i -
i i
1 J
0* "m a'r V
3 1
b E
-1.
E 7
2,I....t....
b
.a D.
5.
10.
15.
20.
TIME ISEC)
MISSISSIPPI FEEDWATER CONTROLLER FAILURE,104.2% P011ER,110.9% FLOW.
FIGURE POWER & LIGHT 320*F TFW EOCl-2K MWD /T 15.B.2-E (Cont.)
15.B-33
i 15.B.3 Stability Analysis
~
General Design Criterion 12 (10CFR50, Appendix A) states that power oscilla-tions which result in exceeding specified acceptable fuel design limits are either not possible or can be readily and reliably detect'ed and suppressed.
Reference 15.B.11-5 provides stability compliance critiera for GE fueled BWRs operating in the vicinity of limit cycles. Analyses in Reference 15.B.11-5 demonstrate that for neutron flux limit cycle oscillations just below the 120%
nantron flux scram setpoint, fuel design limits are not exceeded for those GE BWR fuel designs contained in General Electric Standard Application for Reactor Fuel (GESTAR, Reference 15.B.11-6).
For demonstration of compliance with GDC 12, the generic analyses of Reference 15.B.11-5 are independent of stability margin since the reactor is already assumed to be in limit cycle oscillations (no stability margin).
This implicity covers any variations in s'tability margins caused by FWHOS operation.
The fuel performance during limit cycle oscillations is characteristically dependent on the fuel design and certain system is characteristically dependent on the fuel design and certain system features (e.g., high neutron flux scram setpoint, inlet orifice diameter of channel) and as such it is possible to determine the acceptability of fuel design independent of plant and cycle parameters.
The effects of any changes caused by FWHOS operation (e.g., power distribution inlet subcooling) are covered by the bounding analyses performed in Reference 15.B.11-5.
Therefore, the stability compliance criteria of Reference 15.B.11-5 are satisfied for operation in the FWHOS mode.
l l
[
CS:csc:gc/Il0236*
15.B-34 11/13/84
15.B.4 Loss of Coolant Accident Analysis s
A Loss of Coolant Analysis (LOCA) was performed for GGNS with FWHOS operation.
Reduction of feedwater temperature results in increased subcooling in the vessel thus increasing the mass flow rate out of a LOCA break.
However, an increase in initial total system mass and a delay in lower plenum flashing also occur. They act together to decrease the impact of increased flow out of the recirculation line break.
As a result of this offsetting effect, the peak cladding temperature (PCT) was shown to be lower than the 2098"F value reported for GGNS and below the 2000*F 10CFR50.46 cladding temperature limit.
9 l
CS:csc:gc/Il0236*
15.B-35 11/13/84
15.B.5 Containment Response Analysis The impact of FWH05 on the containment LOCA response was evaluated.
Both Main Steam Line (MSL) break and recirculation line break were analyzed over the entire operation power / Flow region.
Even though the reduced feedwater tempera-ture increases the subcooling of the coolant, the mass flow rate from the postulated recirculation pipe break also increases, but is limited by the critical flow of the break.
The final outcome is that the peak drywell and containment pressures under the FWH05 conditions are bounded by the design values in Chapter 6.
1 l
1 fff{3bgc/Il0236*
15.B-36
i 15'.B.6 Acoustic Load and Flow Induced Loads Impact on Internals Acoustic loads are loads on vessel internals created by a sudden LOCA.
Acous-tic loading is proportional to total pressure wave amplitude in the vessel due to LOCA.
Loadsarecreatedontheshroud,shroudsupportandjetpumpsduetohigh velocity flow in the downcomer in a postulated recirculation line break.
These flow induced loads are affected by the critical mass flux rate out of the break. The reactor internals most impacted by acoustic and flow induced loads l
underFWH05operationaretheshroud,shroudsupportandjetpump. The impact i
on these components were evaluated over the operating power flow region.
FWHOS increases subcooling thus reduces critical flow out of the break.
- However, FWH05 also increases density.
The analyses concluded that these components have enough design margin to handle the loading during FWH05 operation.
CS:csc:gc/Il0236*
15.B-37 11/13/84
j 15.B.7 Feedwater Nozzle Fatigue Usage An evaluation was performed on the feedwater nozzle in GGNS for FWHOS opera-tion. Assuming a full single 18 month cycle operation with feedwater heater out of service based on an 80% capacity factor would result in 438 full power days
)
operation per cycle.
This will result in an additional 0.0214 fatigue usage factor over 40 years of continuous FWHOS operation. An evaluation was also
- performed assuming end of cycle operation with feedwater temperature between 420*F and 250'F for 41 full power days per year for 40 years.
The resultant fatigue usage factor increases by 0.001.
The total fatigue usage factor will still be less than 0.8, which is below the limit of 1.0.
The above assumption of 40 years of continuous FWHOS operation is extremely conservative.
The nozzle fatigue is expected to be much less than the results presented above.
l 1
1 CS:csc:gc/Il0236*
15.B-38 11/13/84 i
15.8.8 Feedwater Sparger Impact Evaluation An evaluation was performed to examine the impact of FWHOS operation on the feedwater sparger for GGNS.
Six cases were analyzed to determine the number of days allowable per year (for 40 years) for FWHOS operation without exceeding the feedwater sparger fatigue usage factor limit of 1.0.
Results of this study is presented in Table 15.B.8-1.
This table indicates that the 40 year average number of days allowable during an operating year for FWHOS operation decreases with lower feedwater temperature and with end of cycle Final Feedwater Tempera-ture Reduction (FFWTR) operation.
In addition, this value is also sensitive to the temperature reduction step used in FFWTR operation.
This reduction in sparger lifetime is mainly of economic concern.
i 1
t l
l l
CS:csc:gc/Il0236*
15.B-39 11/13/84
Table 15.B.8-1 i
~
Summary of Feedwater Sparger Fatigue Analysis Results for FWHOS Operation FFWTR Allowable Number of Days per Year **
To 250'F For FWHOS operation for 40 Years in 41 days At FWT Of for 18-month cycle for 40 years 370'F 320*F No FFWTR 256 61 3 Step
- 127 21 7 Step
- 144 24
- 3 Step FFWTR is s3 - 50'F Step
- 7 Step FFWTR is s7 - 25*F Step
- This evaluation assumes 70% capacity factor.
Allowable number of days which results in a feedwater sparger fatigue usage factor of 1.0.
CS:csc:gc/Il0236*
15.B-40 11/13/84
.=
i 15.B.9 Reactor Protection System Setpoint PNPP's turbine stop and control valve closure scram functions have a low power limit at 40% NB rated power with 420*F feedwater.
A given core power based on 420'F feedwater will not produce the same steam flow as the same core power based on 320'F or 370*F rated feedwater. Turbine steam flow characteristics change when feedwater temperature is reduced. Thus, it is necessary to read-just turbine stop and control valve scram bypass setpoints for FWHOS operation.
I At reactor power levels where significant amounts of steam are being generated, the fast closure of turbine stop or control valves will result in rapid reactor vessel pressurization. When pressure increases, power increases, especially if.
the bypass valves fail to open.
For this reason, scram occurs on turbine stop valve position and control valve fast closure to provide margin to the core thermal-hydraulic safety limit.
At low power levels, high neutron flux scram and vessel pressure scram ~and other normal scram functions are sufficient to provide the safety limit margin even with stop valve or control valve sudden closures. The required lower bound for stop valve position and control valve fast closure scram is 40% of NB rated power.
This is equivalent to $30% of the first-stage pressure (in psia) that would exist at turbine valves wide-open (WO) steam flow conditions. Turbine first-stage pressure is the parameter used to enable the turbine valve closure scram functions. Therefore, below
+40% power, the turbine stop valve or control valve scram functions are not l
enabled.
t As feedwater temperature is reduced, steam flow decreases.
Since the FFWTR process maintains rated core thermal power, the steam flow reduction means that I
the turbine first-stage pressure versus power relationship is altered.
A new setpoint is established for the trip units prior to commencement of each FWHOS operation at each operating cycle. The recommended setpoints for the. turbine j
stop or control valve RPS scram function are 21.0 1 0.5% of calibrated span for i
420'F to 370'F rated feedwater temperature operation and 18.0 1 0.5% of calibrated span for 370'F to 320'F rated feedwater temperature operation.
CS:csc:gc/Il0236*
15.B-41 11/13/84
15.B.10 Miscellaneous Impact Evaluation
(
15.B.10.1 Feedwater System Pipina l
A standard stress analysis was performed on the feedwater system piping up to the first feedwater guide lug outside the containment for feedwater temperature
_at 250*F.
Results of the study show that with the additional FWHOS operations, the feedwater piping fatigue usage factor still meets the allowable limit of 1.0.
15.B.10.2 Impact on Anticipated Transient Without Scram (ATWS)
An impact evaluation was performed which shows that reducing feedwater tempera-ture helps to reduce the consequences of an ATWS event.
The worst ATWS event, MSIVC, was used to evaluate the FWHOS impact. As a result of reduced feedwater i
temperature, steam flow and core average void fraction are reduced.
The lower steam flow rate is produced because more of the core heat is needed to heat up the colder moderator in the core. Therefore, less steam'is generated at its rated power as feedwater temperature is decreased, a case les,s severe than when i
the plant is operating with feedwater temperature at 420*F.
15.8.10.3 Annulus Pressurization load (APL) Impact l
A boundary analysis was performed to determine the impact of FWHOS operation on annulus pressurization loads (APL).
It is found that FWHOS has a small impact l
on annulus pressurization loads. The FWHOS APL is bounded by the normal
/
operation APL limits.
15.B.10.4 Fuel Mechanical Performance l
Evaluations were performed to determine the acceptability of GGNS FWHOS opera-tion on GE-6 fuel rod and assembly thermal / mechanical performance.
Component pressure differential and fuel rod overpower values were determined for antici-pated operational occurrences initiated from FWHOS conditions.
These values were found to be bounded by those applied in the fuel rod and assembly design bases and therefore, GGNS FWHOS operation is acceptable and consistent with l
fuel design basis.
l CS:csc:gc/Il0236*
15.B-42 11/13/84
l 15.B.11 References 15.B.11-1 " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" NED0-24154 Oct. 1978.
15.B.11-2 R. B. Linford " Analytical Methods of Plant Transients E aluations for the General Electric Boiling Water Reactor" NED0-10802 April 1973.
15.B.11-3 "Three Dimensional BWR Core Simulator" NED0-20953-A, January 1977.
15.B.11-4 Letter, J. S. Charnley (GE) to F. J. Miraglia (NRC), " Loss of Feedwater Heating Analysis", July 5, 1983 (MFN-125-83).
.~
15.B.11-5 " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria" NEDE-22277-P, December 1982.
15.B.11-6 " General Electric Standard Application for Reactor Fuel" NEDE-24011-P-A, January 1982 I
i CS:csc:gc/Il0236*
15.B-43 11/13/84
ATTACHMENT I I
PROPOSED TECHNICAL SPECIFlCAT10N MARKUPS FOR 1
FEEDWATER EATER (S) OUT OF SERVICE (FWHOS) l (IN STANDARD OPERATING REGION)
IMPLEENTATION BASES:
APPENDIX 158 0F FSAR
/
1 At-1 1
4 a
POWER DISTRIBUTION LIMITS
(
(
3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION The MINIMUM CRITICAL POWER RATIO (MCPR) sha 3.2.3 ual to or greater than both MCPR, and MCRP,and 3.2.3-2.
limits at indicated core flow en shown in Figures 3.2.3-1 i
g g (*)
e APPLICABILITY: OPERATIONAL CONDITION 1; when THERMAL POWER is greater than or equal to 2 n of RATED THERMAL POWER.
ACTION:
With MCPR 1ess than the applicable MCPR limits determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
~'
1
$URVEILLANCE REOUIREMENTS
\\
4.2.3 MCPR shall be determined to be equal to or greater than the applicable f
j
(
MCPR limits determined from Figures 3.2.3-1 and 3.2.3-2:
(
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and l
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating.'
c.
with a LIMITING CONTROL ROD PA1 TERN for MCPR.
i i
~
l l
of Ned MWkfT AT rder$ *ko NC IArsned yMut, e st>%
4.<mperabre 4srm et rd J -4Ader %peatre (~42e'F (M.
'5 l
sack as. protong,4 y,,novat of f,,A, der bedem) -(m,.
Service.
i
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20 40 60 80 100 120 CORE POWER (5 RATED)
These MCPRp limits assume that the TCV/TSV scram bypass setpoint (see note (h) of Table 3.3.1-1) and E0C-RPT bypass setpoint (see note (b) of Table 3.3.4.2 2) are consistent with corresponding steam flow. For AT>0, the trip setpoint shall be conservatively reduced from <25.4% of calibrated span en increasing turbine first-stage pressure to c15%. The allowable value shall be consistently reduced from
<26.9% of calibrated span to <16.5%.
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ADDENDUM TD
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GGNS MAXIMUM EXTENDED DPERATING DOMAIN ANALYSIS i
This addendum provides supplemental informadion which was 1
requested by the NRC f or clarification of some contents of the i
GGNS Maximum Extended Operating Domain Analysis Report, March 1986.
- 1. Section 15.D.4.1 Abnormal Operating Transients In page 15.D-4-2 it was stated that the results of the Grand Gulf unique evaluation show that the liCPR results for all the cases analyzed are bounded by the bounding BWR/6 Standard Plant analysis presented in Table 15.D.4-1. How-ever, a comparison of the ACPR results in Tab'e 15.D.4-1 and Table 15.D.4-5 shows that this statement is correct except one case; the 104.2/75 FWCF 6CPR O.084 in Table 15.D.4-1 does not bound the 4CPR O.09 of the same tranc.i-ent in Table 15.D.4-5.
The BWR/6 bounding ACPR results in Table 15.D.4-1 are the I
values calculated per ICPR = 1.06 + ACPR f or each tranmi-ent. Therefore, the ICPR values are not the same as those given in Table 15.D.4-5. For the case in question, the ICPR is 1.096 compared to 1.27 for the Grand Gulf unique case.
The bounding BWR/6 A CPR f or ICPR of 1.27 is 0.09 (>0.084) which is equal to the Grand Gulf unique value of 0.09.
Therefore, the Grand Gulf unique ACPR results for all the cases analyzed are either equal to or bounded by the bound-i ing BWR/6 Standared Plant analysis.
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- 2. Section 15.D.6 Loss of Coolant Accident Analys'is The bounding BWR/6 Loss of Coolant Accident (LOCA) analysis was performed in the Maximum Extended Dperating Domain (MEOD) boundary and the results were reviewed for GGNS.
It was concluded that the current MAPLHGR limits presented in Chapter 6 are adequate to cover the entire MEOD. The 1
results of the study are described below.
A generic BWR/6 LOCA analysis was perf ormed along the entire MEOD boundary as defined in Figure 15.D.3-1. The analysis J
was conducted to define MAPLHGR restrictions (multiplines) versus core flow (if any) required to cover operation in the MEOD region. This analysis was performed by comparing the g;
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l values of key parameters which affect the peak cladding tanperature (PCT) along the HEOD houndary to the equivalent values in the GGNS FSAR. Based en this evaluatien, it was concluded that operation in the MEOD region results in no more than 5'F PCT increase over the GGNS FGAR 6.3 ECCS analysis.
Operation at core flows greater than rated tends to reduce the calculated PCT slightly due to the higher coru flow during the period when the recirculation pumps are coasting down. It is the extended load line in the MEOD region which is a ccncern with regard to ECC5 performance and PCT res-pense. The higher rod line will permit a higher power (higher initial storad energy in the fuel) at a given flow.
This increases tha chance of losing nucleate boiling at the highest power axial node prior to the time of Jet punp uncovery. This phenomena is called early boiling'trenmition (BT), and could affect the calculated PCT.
j The two major parameterm that affect PCT in the design basis LOCA calculation which are sensitive to the higher core power and lower ccre flow are the time of BT at the high power. axial node of the limiting fuel assembly and the cal-culated reflooding tion. Early BT results in a less effi-cient removal of the initial stored anergy from the fuel, which tends to increase the calculated PCT. The lower initial pnwer at lower core flow tends to decrease the cal-culated PCT. This occurs because the lower power results in lower core spray vaporitation, leading to less counter current flow limita. tion (CCFL) at the upper tie plate in the fuel bundics and an earlier core reflooding time.
The variation of the bundle inlet flow during a LOCA event is determined by a number of parameters, the most important being the break size, the water inventory in the reactor at the start of the event, and the steady state core poder and flow conditinnu. The first two of these are accounted for in the stundtrd FSAR LOCA analysis. The effect of variations in the third was accounted f or by perf orming analyses at the powar flow points along the MEDD boundary of Figure 15.D.
3-1.
The asraumed intial ninimum critical power ratio (MCPR) is also an important parameter in det ermining whether or not early BT will occur at the high power muial node. Credit was taken for the current tech spec requientaant on MCPR versus j
core flow with.an additional 2 percent conservatism added to satisfy 10CFR50 Appendix K. The required HEOD flow dependent NCPR limit was reviewed to show that there in no significant i
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I RESULTS The generic ECCS analysis f or the BWR/6-218 standard plant showed no early BT prior to jet pump uncovery at the high power node for any of the analysis points on the MEOD boun-dary. Thus, no MAPLHGR reductions are required for operation in the MEDD region from ECCS considerations. The BWR/6-218 was selected as the bounding plant based on lower initial i
and minimum hot bundle mass flux as shown in Table 1.
The lower initial and minimum hot hundle mass flux f or a BWR/6-I 218 makes it the most sensitive BWR/6 to lower initial core flows and thus bounding in terms of any PCT increase which eight result f rom operation in the MEDD. The most limiting
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point in the analysis was found to be at 102 percent power /
85 percent core flow. As stated above, the mass flux in the hot bundle is a key parameter in determining whether or not early BT will. occur. Table 2 shows the effect of the reduced core flow on the time of BT. As can be seen from Table 2, early BT, will not occur at the high power axial node along the MEOD boundary. Tables 1 and 2 also show that the SENS response was very similar to that for the BWR/6-218 plant.
j Thus, the results of the generic study are applicable to SGNS and no MAPLHGR reductions are required f rom ECCS con-siderations. Figure i shows the normalized core flow versus time plot for the GENS DBA recirculation suction line break l
at 100 percent and 85 percent core flow. Figure 2 shows the calculated MCPR versus time f or the BWR/6-218 plant at 100, 1
65, and 75 percent core flow to demonstrate that the most severe response is at B5 percent core flow.
The slightly earlier high power node BT times shown in Table 2 at.75, 80, and 110 percent core flow occur af ter jet pump uncovery f or the BWR/6-218 plant and are estimated to 0
result in a PCT increase of about 2 F. This estimate is based on generic sensitivity studies which show a PCT increase of about 20 F f or a one second change in dryout time when dryout occurs near ten seconds. The changes in dryout time are of similar magnitude for CCND, with a con-servative estimate of less than 5 F PCT increase over the Maximum PCT reported in 6.3 of FSAR.
The ECCS analysis in MEOD described above concluded that the GGNS FSAR Chapter 6 results are impacted by Inss than 5* F PCT and therefore meet the 10CFR50.46 limits.
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- 3. Section 15.D.4.3 Flow Runout Transient i
In this section, it was stated that this event was j
reanalyzed as part of the MEOD program to include the highest rod line for the ELLR, up to 102.5% maximum flow and the ICFR, up to 107% maximum core flow. The following supplements this statement.
The MCPR4 curves for 102.5% and 107% maximum core flow were generated following the same basic procedure and approach i
as used for a BWR/4 MG set plant and therefore, the Tech Spec would be applied in similar manner. The only difference j
is that a scoop tube limits the maximum core flow in the MG j
set plant, whereas in a Flow Control Valve plant like GGNS the electric output of the Flow Controller in the recircu-lation flow control system limits the valve position and maximum core flow. The flow limit is set in such a manner that core flow does not exceed the maximum flow at the rated power.
i The MCPRf calculation is based on a rod line with the limit-ing slope which bounds possible variations of the slope under any Xenon condition, equilibrium or non-equilibrium.
The approved BWR Core Therma 1 hydraulic Analysis Code was used for MCPR calculation. The MCPRf for 102.5% maximum core flow was calculated as follows. First, the MCPR of the peak l
power bundle in the core was put on the saf ety limit MCPR of i
1.06 at 104.2% power (105% steam flow) /102.5% core flow. This
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was done by adjusting the power of the peak power bundle.
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Next, the peak power bundle MCPR was calculated along the I
limiting rod line to determine the MCFRf as function of core flow. A similar procedure was used for the 107% maximum core j
flow MCPR4 calculation. The only difference was that the peak power bundle MCPR was put on 1.06 at 107% core flow and core power corresponding to 107% core flow on the same rod
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The slow flow runout analyses have been performed for the steepest rod line in the ELLR region and compared to results of the analysis for the 105% rod line. The evaluation has shown that the MCPRp curves based on the 105% rod line j
bound the highest rod line came (ELLR case) because the maximum core flow attainable on the ELLR rod line is less for a given maximum recirculation FCV position due to higher two-phase pressure drop.
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TABLE 1 Effect of Core Flow on Hot Bundle Inlet Mass Flux Initial Initial Hot Bundle Minimum Hot Bundle Core Flow Mass Flux Mass Flux *
(X Rated)
(1bm/hr-ft2)
(Ibm /hr-ft2)
BWR/6-21B Std Plant 110 1.177 E6 0.463 E6 100 1.050 E6 0.411 E6 85 0.890 E6 0.363 E6 BO O.835 E6 0.341 E6 75 0.782 E6 0.319 E6 70 0.731 E6 0.298 E6 65 0.681 E6 0.275 E6 60 0.632 E6 0.248 E6 55 0.583 E6 0.221 E6 45 0.487 E6 0.157 E6 GGNS 100 1.083 0.424 E6 B5 0.910 0.368 E6
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TABLE 2 Effect of Core Flow on Boiling Transition (BT) in the Hot Channel for DBA Recirculation Suction Line Break Initial Time of BT Time of87 Elevation Core for High for. Upper of Upper Flow Initial Power Node Node Node (X Rated)
MCPR (Sec)**
(Sec)
(Ft from TAF)
BWR/6-218 Std Plant 110 1.147 9.BB*
O.98 2.344 100 1.147 9.96 1.18 4.427 85 1.147 10.10 1.02 4.427 BO.
1.206 9.88*
O.94 2.344 75 1.245 9.92*
9.92 O.
70 1.279 10.16 10.16 O.
65 1.319 10.22 10.22 O.
60 1.343 10.12 10.12 O.
55 1.377 10.62 10.62 O.
45 1.436 10.20 10.20 O.
GGNS 100 1.147 9.60 1.32 4.427 85 1.147 9.54*
1.00 4.427
- Slightly earlier than base case, but PCT impact is less than 5* F.
- BT occurs after jet pump uncovery, which is approximately B seconds after LOCA.
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