ML19296C617
| ML19296C617 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/1980 |
| From: | Engel R, Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19296C611 | List: |
| References | |
| 80NEDO253, NEDO-24235, NUDOCS 8002260715 | |
| Download: ML19296C617 (39) | |
Text
- . 6 80 E 2 3 JANUARY 1980 h
SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 3 P]LR NEula_
u uva260 Y T GENER AL h ELECTRIC
NEDO-24235 80NED253 Class I January 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 3 e
Prepared:
.[L. Rash Senior Engineer Fuel and Services Licensing Approved:
R.E.Engel[ Manager Reloed Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CA LIFORNI A 95125 GEN ER AL h ELE' J TRIC
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULoY This report was prepared by General Electric solely for Carolina Power and Light Company (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending CP&L's operating license of the Brunswick Steam Electric Plant Unit 1. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the Fuel Contract Supplemental Agreement between Carolina Power and Light Company and General Electric Company for Brunswick 1 & 2 dated January 28, 1974, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any repre-sentation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
NED0-24235
- 1. PLANT-UNIQUE ITEMS (1.0)*
Supplemental Transient and GETAB Analyses: Appendix A Transient Analysis Initial Conditions: Appendix B Recirculation Pump Trip: Appendix C New Bundle Loading Error Analyses Procedures: Appendix D Linear lleat Generation Rate for Bundle Loading Error: Appendix E Densification Power Spiking: Appendix F
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0)*
Fuel Type Number Number Drilled Irradiated Initial Core Type 1 64 64 Initial Core Type 3 88 88 Replacement 7DB230 4 4 Reload 1 8DB274L 100 100 Reload 1 8DB27411 40 40 Reload 2 8DRB26511 64 64 Reload 2 8DRB283 68 68 New Reload 3 P8DRB26511 132 132 Total 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 12,755 mwd /t Assumed reload cycle exposure: 14,943 mwd /t Core loading pattern: Figure 1
- ( ) refers to areas of discussion in NEDE-24011-P-A, " Generic Reload Fuel Application," August 1979.
1
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTil - NO VOIDS , 20 C ( 3. 3. 2.1.1 and J . 3. 2.1.2)
BOC k gg Uncontrolled 1.097 Fully Controlled 0.939 Strongest Control Rod Out 0.976 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, Ak 0.007
- 5. STANDBY LIQUID CONTROL SYSTEM SilUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) ppm (20 C, Xenon Free) 600 0.0465
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)
EOC4 Void Coefficient N/A* (c/% Rg) -8.80/-10.99 Void Fraction (%) 41.76 Doppler Coefficient N/A (c/% F) -0.227/-0.215 Average Fuel Temperature (*F) 1356 Scram Worth N/A ($) -38.88/-31.10 Scram Reactivity Figure 2 EN = Nuclear Input Data A = Used in Transicut Analysis 2
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 7x7 8x8 8x8R/P8x8R EOC4 EOC4 EOC4 Peaking Factors 1.24/1.30/1.40 1.22/1.45/1.40 1.20/1.58/1.40 (local, radial and axial)
R-Factor 1.100 1.098 1.051 Bundle Power 5.548 6.184 6.722 (MWt)
Bundle Flow 125.0 112.6 113.4 (103 lb/hr)
Initial MCPR 1.18 1.20 1.20
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Recirculation Pump Trip
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Power Flow $ Q/A s1 v t.CPR Transient Exposure 8x0R/ Plant
(%) (%) (%) (%) (psig) (psig) 7x7 8x8 P8x8R Response Turbine BOC-EOC 104 100 189 107 1171 1201 0.07 0.11 0.11 Figure 3 Trip w/o Bypass
! ass of BOC-EOC 104 100 124 122 <1100 <1100 'O.12 0.14 0.14 Figure 4 Feedwater Heating Feedwater BOC-EOC 104 100 118 110 1045 1081 0.05 0.06 0.07 Figure 5 Controller Failure 3
- 10. LOCAL ROD WITilDRAWAL ERROR (WITil LIMITING INSTR 141ENT FAILURE)
TRANSIENT Sl&SIARY (5.2.1)
Rod MIUCR (kNIIII Position 8x8R/ 8x6R/ Limiting Rod Block (Feet t.CPR Reading Withdrawn) 7x7 8x8 P8x8R 7x7 8x8 P8x8K Rad Pattern 104 3.5 0.12 0.17 0.09 16.5 15.1 16.0 Figure 6 105 4.0 0.14 0.21 0.11 17.3 16.1 17.3 Figure 6 106* 4.0 0.14 0.21 0.11 17.3 16.1 17.3 Figure 6 107 4.5 0.16 0.24 0.13 17.5 16.5 17.7 Figure 6 108 5.0 0.18 0.27 0.16 17.7 16.8 17.9 Figure 6 109 5.0 0.18 0.27 0.16 17.7 16.8 17.9 Figure 6
- 11. OPERATING MCPR LIMIT (5.2)
BOC4 - EOC4 1.21 (7x7 fuel) 1.28 (8x8 fuet) 1.21 (8x8R fuel /P8x8R fuel)
- 12. OVERPRESSURIZATION ANALYSIS SlBDIARY (5.3)
Power Core Flow s1 y Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 104 100 1238 1268 Figure 7 (Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 Reactor Core Stability:
Decay Ratio, x2 / x0 0.75 (105% Rod Line -
Natural Circulation Power)
Channel liydrodynamic Performance Decay Ratio, x2 / xo Extrapolated Rod Block Line - Natural Circulation Power) 7x7/8x8/8x8R channel 0.16/0.28/0.21
- Indicates set point selected 4
NEDO-24235 14 LOSS-OF-COOLANT ACCIDENT RESULTS, (5.5.2)
P3DRB265H Exposure MAPLHGR PCT Location Oxidation (mwd /t) (kW/ft) (
- F) Fraction 200 11.5 2138 0.028 1000 11.6 2146 0.028 5000 11.9 2174 0.030 10,000 12.1 2187 0.031 15,000 12.1 2196 0.032 20,000 11.9 2177 0.030 25,000 11.3 2101 0.024 30,000 10.7 2016 0.018
- 15. LOADING ERROR RESULTS (5.5.4)
Limiting Event: Rotated P8DRB265H MCPR: 1.07
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Function: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 5
NED0-24235
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l llED8gm'G9035Gy8 s 8 8 E][D'8 8 I l l l l l ll IIII 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 FUEL TYPE A = INITIAL CORE TYPE 1 E = RELOAD 1 SDB274L 8 = NOT USED F = R E LOAD 1 SDB274H C = INITI AL CORE TYPE 3 G= RELOAD 2 80R8265H D = REPLACENIENT 7D8230 H =
RELOAD 2 80R8283 J = RE LOAD 3 P80RB265H Figure 1. Reference Core Loading Pattern 6
NEDO-24235 1m #8 C
M -
_ g C - 678 CRD IN PERCENT g _ 1 - NOMINAL SCRAM CURVE IN (-$)
2 - SCRAM CURVE USED IN ANALYSIS -
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Figure 2. Scram Reactivity and Control Rod Drive Specifications 7
l i NEUTRON # LUX 1 VESSEL PfES RISE (PSI) 2AVESLff4:E r1 EAT FLUX 2 SAFETT WLVE FLOW
- 3 CORE INLd7 FLOW 300- 3 FELJEF WQ MF CW_
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Figure 3. Plant Response to Turbine Trip Without Bypass
! NEUTPON HLUX 1 VESSEL PIES RISE (PSI) 2 AVE SURF C E HEAT FLUX 2 RELIEF VLLVE FLOW t e;c, / 3 CCPF It.U 81 FLCd 375' 3 BTPASS vhtVE FLOW 4 CCrt INLt T SUB 4
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Figure 4. Plant Response to Loss of 100*F Feedwater Heating
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TIME (SEC1 TIME (SEC1 Figure 5. Plant Response to Feedwater Controller Failure
NED0-24235 02 06 10 14 18 22 26 51 10 12 47 43 2 6 8 39 35 10 6 26 0 31 36 27 12 2 12 26 Notes: 1. Rod Pattern Is 1/4 Core Mirror Symmetric, Upper Left Quadrant Shown on Map.
- 2. Numbers Indicate Number of Notches Withdrawn out of 48. Blank Is a Withdrawn Rod.
- 3. Error Rod Is 26-35.
Figure 6. Limiting RWE Rod Pattern 11
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NEDO-24235 1.2 ULTIMATE PERFORMANCE LIMIT 1.0 -
NATURAL CIRCULATION o.s -
o 2!
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e 106% ROD LINE
$ os -
8 0.4 -
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o ! I I O 20 40 00 80 too PERCENT POWER Figure 8. Decay Ratio 13
-5 IN 1/DE LTA degrees C (MILLIONTHS) e
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F UE L TE MPER A71. %". I' ;)
Figure 9. Doppler Reactivity Coefficient Comparison for RDA 14
NEDO-24235 200 160 -
G O E O
@ 120 b
N 2
7, 80 x
O BOUNDING VALUE FOR 280 cal /g 4o _ O CALCULATED VALUE i
of I I l o 4 8 12 is ROD POSITION (ft OUT)
Figure 10. RDA Reactivity Shape Function at 20 C 15
NED0-24235 14 12 -
O 10 -
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4 8 -
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O BOUNDING VALUE FOR 290 cal /g 1
I O CALCULATED VALUE 2 -
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ROD POSITION (ft OUT)
Figure 11. RDA Reactivity Shape Function at 286 C 16
NEDO-24235 70 CD -
O BOUNDING VALUE FOR 280 cal/g
~
O CALCULATED VALUE a
i 3 40 -
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Figure 12. RDA Scram Reactivity Function at 20 C 17
NED0-24235 100 -
80 -
E O SOUNDING VALUE FOR 200 cal /g E O CALCULATED VALUE 5
3 E
b
%t 60 N
s C
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40 -
20 -
0" 6 8 0 2 4 ELAPSED TIME (sec)
Figure 13. RDA Scram Reactivity Function at 286*C 18
NEDO-24235 APPENDIX A SUPPLEMENTAL TRANSIENT AND GETAB ANALYSES The analyses in this appendix were performed with the ODYN code.
A-1/A-2
Table L-1 Power Flow & Q/A sl v 8x8R/ Plant Transient Exposure (%) (%) (%) (%) (psig) (psig) 7x7 8x8 P8x8R Response Turbine Trip EOC 104 100 373 117 1190 1215 0.09 0.12 0.14 Figure A-1 No Bypass Feedwater EOC 104 100 114 110 1057 1090 0.04 0.06 0.05 Figure A-2 Controller Failure
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Figure A-2. Plant Response To Feedwater Controller Failure
NEDO-24235 APPENDIX B TRANSIENT ANALYSIS INITIAL CONDITIONS S/RV Lowest Setpoint (psig) 1105 + 1%
S/RV Capacity (%) 87.4 Turbine Pressure (psig) 975.1 B-1/B-2
o NEDO-24235 APPENDIX C RECIRCULATION PUMP TRIP FEATURE C.1 INTRODUCTION Signiricant improvement in thermal margin can be realized if the severity of core-wide pressurization transients is reduced. The Recirculation Pump Trip (RPT) feature accomplishes this by rapidly cutting off power to the recirculation pump motors any time turbine control valve or turbine stop valve fast closure occurs. This results in a rapid reduction in recircu-lation flow and increases the cora void content during the core-wide pressurization transients, thereby reducing the peak transient power and heat flux. A more detailed discussion of the effect of RPT is included in Sect ion C-2.
Basically, the RPT system consists of pressure switches
- installed in the turbine control valves and the position switches
- in turbine stop valves.
When these valves close, redundant breakers between the motor generator sets and the recirculation pump motors are tripped; this releases the recircula-tion pumps to coast down under their inertia. Adding RPT will result in a reduction in CPR for transients involving stop valve or control valve closures.
C.2 EFFECT OF RPT ON PLANT PERFORMANCE C.2.1 Dynamic Characteristics An inherent design characteristic of the boiling water reactor (BWR) is the relationship of the core average moderator density to neutron moderation, which is represented by a negative void reactivity coefficient. This nega-tive void reactivity coef ficient permits load following through control of
- These closureareor the stopsame valveswitches closure.which initiate scram on contcol valve fast By using the same signal to initiate RPT, the necessary hardware modifications are minimized and the scram trip and RPT are initiated simultaneously.
C-1
NED0-24235 the recirculation flow without control rod movement. To increase power, core flow is increased, which decreases the void f raction and increases the neutron moderation and teactor power.
The negative void reactivity characteristic of the BWR dictates the neces-sity for reactivity control during certain operational pressurization events.
The two most limiting events analyzed in a typical plant safety analysis are the rapid turbine stop valve closure (turbine trip) or control valve closure (generator load rejection) with assumed bypass f ailure. In these events, the dome pressure increases rapidly, causing a reduction in the core average void f raction, which increases moderation and results in a positive power increase. This is reflected in decreased margins to pressure and thermal limits.
The physical phenomenon which causes the reduced margins is that the void reactivity feedback, which is due to the pressurization, momentarily can add positive reactivit y to the system f aster than the control rods add negative scram reactivity.
The BWR design provides a system for which reactivity changes have an inverse relationship to the steam void content in the moderator. This void feedback effect is one of the inherent safety features of the BWR system. Any system input which increases reactor power (either in a local or gross sense) pro-duces additional steam voids, which reduces the reactivity and thereby reduces the power. The void f eedback mechanism contributes to the stable regulation of core reactivity and permits load following without use of control rods by varying the recirculation flow. The practical constraints on the void coef-ficient are that it must be large enough to prevent power oscillation due to to not spatial xenon changes yet small enough that pressurization transients unduly limit plant or ration.
The basic phenomenon associated with void feedback is the decrease in neutron moderation resulting f rom an increase in void frac . ion. A spectral shift in the neutron flux occurs wherein the thermal flux, at.d hence the fission rate, decreases and the epithermal flux, and hence the <es- ace capture rate, increases. Conversely, a decrease in void f raction -~res an increase in C-2
NED0-24235 reactivity. ihe void coefficient is predominantly the function of three variables for any fixed bundle geometry: (1) the average volds; (2) enrich-ment; and (3) expesure. As each of these three parameters increases, the absolute magnitude of the void coef ficient increases and becomes more negative.
For pressurization transients, the rate of flux rise is dependent on the magnitude of the void coefficient. The more negative the void coefficient, the greater the flux rise rate. The rate at which the negative reactivity can be added to the core by the scram determines the severity of the transient.
The scram reactivity depends on the ability of the control rods to be in the high flux regions of the core. The minimum scram reactivity occurs at end of cycle (EOC) when control rods are fully withdrawn from the core. In this situation, it takes a longer time for the control rod to travel to a high importance region in the core. For this reason, the pressurization tranciente are most severe near the end of the cycle.
The degree to which the pressure and thermal margins are reduced during pres-surization events depends on the tradeoff between the negative scram and positive void reactivities. Typically, at beginning of cycle (BOC), control rods are partially inserted; this permits a prompt shutdown of the system without a significant decrease in margins. As the fuel cycle proceeds toward EOC, the control rods are withdrawn until, ideally, they are all withdrawn.
Hence, the effectiveness of scram reactivity for shutdown of certain pres-surization transients is decreased as the core approaches EOC conditions.
As discussed above, margins are decreased when the positive void reactivity feedback is inserted at a rate faster than the negative scram reactivity.
Analyses have shown that the transient severity can be signifienntly reduced by a rapid reduction in core flow. This increases the core vo. fraction during pressurization transients and consequently minimizes the power rise experienced. The rapid reduction in core flow necessary to accomplish this effect can be achieved by the prompt tripping of both recirculation pumps.
C-3
NED0-24235 C.3 The RPT system described in Section )N<f has been developed to accomplish this goal.
C.2.2 Thermal Limits Consideration One of the operating fuel thermal limits, the minimum critical power ratio (MCPR), is established such that the most severe abnormal operational transient is not expected to subject more than 0.1% of the fuel rods to boil-ing transition. This is known as the General Electric Thermal Analysis Basis (GETAB). GETAB statistically correlates a calculated MCPR as the con-dition at which less than 0.1% of the fuel rods are expected to experience boiling transition. This value is incorporated into the plant technical specifications as the fuel cladding integrity saf ety limit. An operating limit MCPR is established such that the most severe abnormal operational transient will not result in violating the safety limit. The difference between the actual plant operating critical power ratio (CPR) and the oper-ating limit MCPR is a measure of the thermal margin.
If the normal operating CPR at the licensed power level cannot be maintained above the operating limit MCPR, a plant dera+e will be imposed to assure that the resultant change in CPR from a worst-case abnormal operational transient will not decrease the MCPR below the safety limit. A reduction in severity of the worst transient allows a reduction in the operating limit. Usually either a turbine or generator trip without bypass is the limiting thermal event near EOC. The RPT system is intended to provide improved thermal mar-gin for these limiting events.
C.2.3 Overpressure Protection Considerations In addition to the effect on thermal margins, RPT also has an effect on the overpressure protection margins. There are two types of pressure limits that apply to BWR's. The first pressure limitation is the ASME vessel overprotec-tion limit, which limits the peak vessel pressures to less than 110% of the vessel design limit (1375 psig). Compliance to the vessel design pressure limit is demonstrated by an analysis of the main steam isolation valve (MSIV) closure with indirect scram event (conservatively neglecting the direct scram C-4
NED0-24235 from position switches on the isolation valves). This margin is met by installation of an appropriate number of safety / relief valves. The RPT system has no effect on this analysis because it is not initiated during this event.
Another GE criterion is that associated with unpiped safety valves. In order to preclude steam from being blown directly into the containment, GE recommends that there be adequate margin to the lowest setpoint of any unpiped spring safety valve. This applies to expected operating transients with credit taken for direct scram. The installation of RPT on plants which incorporate recirculation pump trip for anticipated transients without scram purposes will increase this margin.
C.3 GENERALIZED RECIRCULATION PUMP TRIP DESCRIPTION C.3.1 System Function The RPT system, which is designed to improve fuel thermal margin, trips both recirculation pumps upon sensing stop valve closure or f ast control valve closure. The reduced core flow reduces the void collapse in the core during two of the most limiting pressurization events (i.e., turbine and generator trips). Tripping of the recirculation pumps results in a smaller net posi-Live void reactivity addition to the system during these pressurization events. This results in a lower power increase and consequently a lower operating MCPR limit. Although the reduction in core flow in itself may cause a slight decrease in thermal margins, the effect of reduced flow on the power increase is a considerably more dominant effect and the net result is to reduce the thermal severity of the event.
In order f or the RPT system to ef fectively counteract the void collapse effects from pressurization transients, the pump trip must occur very soon after the turbine / generator trip, and the pumps must coast down at a relatively fast rate. If the pump trip and coastdown do not occur quickly, the positive void reactivity feedback caused by the pressurization effects will dominate the transient and no margin improvement will be seen from tripping of the pumps.
C-5
NEDE-24235 Analyses have been performed which demonstrate that the RPT system is made most effective by installing and tripping a line breaker between the recir-culation pump drive motor / generator and the pump motor. Although a motor /
3enerator field breaker trip has cost advantages over a line breaker, the response characteristics from such a trip do not achieve significant improve-n ents in thermal margins. Upon tripping the field breakers, the drive motor generator continues to momentarily supply some reduced power to the pump motor due to the time required for the generator field and line current to drop to zero. This results in reduced effectiveness of the system.
In order to achieve the desired improvements in thermal margins for the turbine / generator trips, the supply current to the pump motor must be termi-nated quickly after receipt of the signal from the switches in the turbine stop valves or in the turbine control volves. The line breaker pump trip does achieve the desired system goal.
C.3.2 Summary System Description The RPT system includes all equipment that trips recirculation pump motors from their power supplies in response to a turbine / generator trip or load rejection. The RPT system is designed to be of quality consistent with the reactor protection system functions which provide protection for the same events. The system consists of turbine control and stop valve closure sen-sors, separate division logic and two circuit breakers for each pump motor.
The RPT system is designed to be operable whenever the turbine generator trip scram is operable (i.e., above approximately 30% reactor thermal pressure).
Existing turbine first-stage pressure sensors will prevent RPT initiation for turbine-generator trips occurring below the existing 30% power bypass of turbine and generator trip scram signals.
The RPT system design includes two separate trip divisions with each having two separate trip channels, sensors and associated equipment for each meas-ured vuriable. The system is designed to meet the single-failure criterion such that any single trip channel (sensor and associated equipment) or system component failure shall not prevent the system from performing its intended C-6
NEDO-24235 safety function. Electromechanical relays used as the logic elements within the system and the system logic are of the failsafe type (i.e., trip on loss of electrical power).
The RPT system is designed to accomplish the desired function and to minimize the effect of this additional system on plant availability. The system logic is designed such that it will not cause the inadvertent trip of more than one pump given a single component failure in the system. Each trip division is clearly identified to reduce the possibility of inadvertent trip of the recirculation pump during routine maintenance and test operations. Redundant sensor circuits in each division (sensors, wiring, transmitter, amplifiers, etc.) are electrically, mechanically, and physically independent so that they are unlikely to be disabled by a common cause except for an electrical power failure.
Capability is provided for testing and calibration of the system logic and circuit breakers. Provisions are made to allow closure of stop valve and fast closure of turbine control valve separately at least one valve at a time (for normal routine valve test purposes) without causing a pump motor trip. The system input sensors and the division logic are capable of being checked one channel or division at a time. The sensors and system logic test or calibra-tion during power operation will not initiate pump trip action at the system level.
C-7/C-8
I;EDO-24235 APPENDIX D NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES The bundle loading error analyses results presented in Section 15 in this supplement are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error events. The use of these new anal-ys;3 procedures is discussed below.
NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error analysis results presented in this supple-ment are based on the new analysis procedure described and approved in Ref-erence D-1. This new method of performing the analysis is based on a more accurate detailed analytical model.
The principal difference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, causes the bundle to lean. The effect of the variable water gap is to reduce the power peaking and the R-factor in the upper regions of the limiting fuel rod. This results in the calculation of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.
NEW ANALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analyses results presented in this supplement are based on the new analysis procedure described in Reference D-1.
This new method of performing the analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core.
D-1
NEDO-24235 The use of the statistically corrected Haling analyses procedure indicates that the minimum CPR f or mislocated bundles is greater than the safety limit (1.07) for all exposures throughout the cycle.
REFERENCES D-1. Safety Evaluation Report (letter) D.G. Eisenhut (NRC) to R.E. Engel (GE),
MFN-200-78, dated May 8, 1978.
D-2
NEDO-24235 APPENDIX E LINEAR HEAT GENERATION RATE FOR BUNDLE LOADING ERROR 16,8 kW/ft E-1/E-2
NED0-24235 APPENDIX F DENSIFICATION POWER SPIKING Reference F-1 documents the NRC staff position that ". . . It (is) acceptable to remove the 8x8 and 8x8R spiking penalty f actor f rom the plant Technical Specification for those operating BWR's for which it can be shown that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's."
The BSEP-2 Reload-3 submittal contains the required information to demon-strate that the stated criterion is met for BSEP-2, Reload 3. Section 10, Rod Withdrawal Error, and Appendix E (Linear Heat Generation Rate for Bundle Loading Error) include the densification ef fect in the calculated LHGR of the 8x8 fuels.
REFERENCE F-1 " Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking Due to Densification Effects in BWR 8x8 Fuel Design and Performance," Reactor Safety Branch, DOR, May 1978.
F-1/F-2
a wE .
GEN ER AL h ELECTRIC