ML19259B414
| ML19259B414 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/31/1979 |
| From: | Brugge R, Ervin A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19259B412 | List: |
| References | |
| NEDO-24587, NUDOCS 7902090266 | |
| Download: ML19259B414 (41) | |
Text
'
i.
"'75s's" JANUARY 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 2 (RECIRCULATION PUMP TRIP FEATURE) 1 0
02G6 GENER AL $ ELECTRI
NEDO-24587 Class I January 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 2 (Recirculation Pump Trip Feature)
}
S/
Prepared:
A. M. Ervin, Engineer Operating Licenses II Approved:
p R. O.
- ugge, anager 0,serating Licenses II NUCLE AR ENERGY PROJECTS DIVISION + GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL h ELECTRIC
NEDO-24587 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Carolina Power and Light (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Commission (USNRC)
Company for amending CP&L's operating license of the Brunswick Steam Electric Plant Unit 2.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or pro-vided 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 contract between Carolina Power and Light Company and General Electric Company for nuclear fuel and related services for the nuclear system for Brunswick Steam Electric Plant, dated January 26, 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 tb.t 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 representation 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.
NEDO-24587 1.
PLANT UNIQUE ITEMS (1.0)*
Rotated Bundle Analysis Procedure: Appendix A Total Number and Capacity of Safety / Relief Valves: Reference 2 Fuel Loading Error LHCR: Appendix B ODYN Transient Calculation Results:
Appendix C Recirculation Pump Trip Feature:
Appendix D 2.
RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0)
Fuel Type Number Number Drilled Irradiated Initial Core Type 1 108 108 Initial Core Type 3 176 176 7DB230 4
4 8DB274L 100 100 8DB274H 40 40 New 8DRB265H 64 64 8DRB283 68 68 Total 560 560 3
REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure:
10,938 mwd /t Assumed reload cycle exposure:
13,200 mwd /t Core loading pattern:
Figure 1 4.
CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM 0
WORTH - NO VOIDS, 20 C ( 3.3.2.1.1 and 3.3.2.1.2 )
BOC k 77 Uncontrolled 1.119 Fully Controlled 0.958 Strongest Control Rod Out 0.989 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, Ak 0.000 5.
STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) 0 ppm (20 C, Xenon Free) 600 0.033 a( ) refers to areas of discussion in Reference 1.
1
NEDO-24587 6.
RELOAD UNIQUE TRANSIENT ANALYSIS INPUTS (3 3 2.1.5 and 5.2)
E00 Void Coefficient N/Ae (t/% Reg) 8.10/10.13 Void Fraction (%)
41.76 Doppler Coefficient N/A (t/%0F) 0.1938/0.1841 Average Fuel Temperature ( F) 1538 Scram Worth N/ A ($)
38.85/31.08 Scram Reactivity vs Time Figure 2 7.
RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITI AL CONDITION PARAMETERS (5.2)
Exposure EOC EOC EOC Peaking factors (local, radial and axial) 1.24/1.284/1.40 1.22/1.448/1.40 1/20/1.585/1.40 R-Factor 1.100 1.098 1.051 Bundle Power (MWt) 5.481 6.17 5 6.753 Bundle Flow (103 lb/hr) 124.5 113 0 1 14.0 Initial MCPR 1.20 1.20 1.20 8.
SELECTED MAR 3IN IMPROVEMENT OPTIONS (5.2.2)
Recirculation Pump Trip eN = Nuclear Input Data A = Used in Transient Analysis 2
CORE-WIDE TR ANSIENT ANALYSIS RESULTS (5.2.1)
Power Flow Q/A P31 P
ACPR Plant V
Transient Exposure
(%)
(%)
(%)
(%)
(psig) _(psig) 7x7 8x8/8x8R Response Load Rejection without Bypass EOC3 104 100 138.8 100 1168 1199 0.03 0.06 Figure 3 Loss of 100 F Feedwater Heating 104 100 117.6 115.8 1019 1068 0.11 0.13 Figure 4 Feedwater Controller Failure E0C3 1 04 100 109.4 104.9 1028 1073 0.05 0.06 Figure 5 10.
LOCAL ROD WITHDR AWAL ERROR (WITH LIM 1 TIN 3 INSTRUMENT FAILURE)
SUMMARY
(5.2.1)
Rod Position Rod Block (Feet ACPR MLHGR (kW/f t)
Limiting Reading?
Withdrawn) 7x7 8x8 8x8R 7x7 8x8 6x8R Rod Pattern 104 35 0.11 0.08 0.16 16.25 13.85 11.65 Figure 6 105n 4.0 0.14 0.09 0.19 17.45 14.85 12.10 Figure 6 106 4.5 0.16 0.11 0.22 18.40 15.70 12.85 Figure 6 107 5.0 0.18 0.12 0.25 19.10 16.40 13.45 Figure 6 108 5.5 0.19 0.14 0.27 19.10 16.40 13.45 Figure 6 109 6.0 0.20 0.15 0.29 19.55 16.95 13.90 Figure 6 11.
OPERATING MCPR LIMIT (5.2)
BOC3 - EOC3 1.26 (8x8/8x8R fuel) 1.21 (7x7 fuel) 12.
OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Ps1 P
Plant Power Core Flow v
(%)
(%)
(psig)
(psig)
Response
MSIV Closure (Flux Scram) 104 100 1218 1263 Figure 7
' Indicates setpoint selected 3
NEDO-24587 13.
STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio:
Figure 8 Reactor Core Stability:
Decay Ratio, x2/XO 0.62 (105% Rod Line - Natural Circulation Power)
Channel Hydrodynamic Performance
/
Decay Ratio, x2 x0 (105% Kod Line - Natural Circulation Power) 8x8/8x8R channel 0.28 7x7 channel 0.13 14.
, LOSS-OF-COOLANT ACCIDENT RESULTS, (5.5.2) 8DRB265 Exposure MAPLHGR PCT Local Oxidation (mwd /t)
(kW/ft)
(OF)
Fraction 200 11.5 2154 0.030 1,000 11.6 2156 0.029 5,000 11.9 2192 0.032 10,000 12.0 2196 0.032 15,000 12.0 2200 0.033 20,000 11.8 2197 0.033 25,000 11 3 2138 0.027 30,000 10.7 2056 0.021 8DRB283 Exposure MAPLHGR PCT Local 0xidation (mwd /t)
(kW/ft)
(*F)
Fraction 200 11.2 2122 0.027 1,000 11.2 2117 0.026 5,000 11.8 2184 0.032 10,000 12.0 2197 0.033 15,000 11.9 2194 0.032 20,000 11.8 2197 0.033 25,000 11.3 2132 0.027 30,000 11.1 2106 0.025 4
NEDo-24587 15.
LOADING ERROR RESULTS* (5.5.4, Appendix A)
Limiting Event:
Rotated Bur 11e 8DRB283H or 8DRB265H MCPR:
1.07**
16.
CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient:
Figure 9 Accident Reactivity Shape Functions:
Figures 10 and 11 Scran Reactivity Functions:
Figures 12 and 13
- Using New Rotated Bundle Analysis Procedures described in Appendix A
- Includes added penalty of 0.02 imposed by NRC, 5
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- 80274L Figure 1.
Reference Core Loading Pattern 6
XEDO-24587 100 45 C-678 CRO IN PERCENT 1 NOMIN AL SCR AM CURVE IN (-3) 90 2 SCRAM CURVE USED IN ANALYSIS 40 80 36 70 30 60 25 E
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Scram Reactivity and Control Rod Drive Specifications 7
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NEDO-24587 02 06 10 14 18 22 26 30 51 47 43 34 6
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23 NOTES 1.
ROD PATTERN IS 1/4 CORE M:RROR SYP.".?ETRIC UPPER LEFT QUADR ANT SHr/WN ON M AP 2.
NUMBERS INDICATE NU'.iBER OF NOTCHES WITHDR AWN OUT OF 48 BLANK IF A WITHDR AWN ROD 3 ERROR ROD IS (22.29)
Figure 6.
Limiting RWE Rod Pattern 11
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NEDO-24587 1.2 ULTIMAT 1.0 - - - - - - - -
E ST ABILITY LIMIT 08 x
N 5
9 0.6 c
><Uy N ATUR AL CIRCUL ATION O4 105% ROD LINE 0.2 0.0 0
20 40 60 80 100 POWLR (%)
Figure 8.
Decay Ratio 13
O BOUNDING VALUE FOR 280 CAL /G COLD 0 BOUNDING VALUE FOR 280 CAL /G HSB O -
6 CALCULATED VALUE - COLD Q CALCULATED V ALUE - HSB -8 e
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400 800 1200 1600 2000 2400 FUE L TEMPER ATURE ('Cl Figure 9.
Doppler Reactivity Coefficient Comparison for RDA 14
NEDO-24587 28 O BOUNDING VALUE FOR 280 CAL /G O CALCULATED VALUE 24 20 G
5Oz 16 S
O E_
xn i
2 h
12 O
Ew" 1
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RDA Reactivity Shape Function at 20*C 15
NEDO-24587 28 24 20 G
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O BOUNDING VALUE FOR 280 CAL /G d
O CALCULATED VALUE I
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Figure 11.
RDA Reactivity Shape Function at 286*C 16
NEDO-24587 90 0 BOUNDING VALUE FOR 280 CAL /G O CALCULATED VALUE 80 70 60 G
ko k
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Figure 12.
RDA Scram Reactivity Function at 20*C 17
NEDO-24587 140 O BOUNDING VALUE FOR 280 CAL /G O CALCULATED V ALUE 120 -
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Figure 13.
RDA Scram Reactivity Function at 286*C 18
NEDO-24587 REFERENCES 1.
" General Electric Boiling Water Generic Fuel Application," NEDE-240ll-P, Revision 3, March 1978.
2.
Letter No. NG-77-1060 from E. E. Utley (CP&L) to A. Schwencer (NRC),
September 20, 1977.
19/20
NEDO-24587 APPENDIX A NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES The bundle loading error analyses results presented in Section 15 in this supplement are cased on new analyses procedures for bo 5 the rotated bundle and the mislocated bundle loading error events.
The use of these new analyses pro-cedures is discussed below.
A.1 NEW ANALYSES PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error event analyses results presented in this sup-plement are based on the new analyses procedure described in References A-1 and A-2.
This new method of performing the analyses is based on a more detailed analysis model, which reflects more accurate analyses than that used in previous analyses of this event.
The principle difference between the previous analyses procedure and the new analyses procedure is the modeling of the water gap along the axial length of the bundle. The previous analyses used a unif orn, water gap, whereas the new analyses utilize a variable water gap which is representative of the actual condition.
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 cal-culation of a reduced ACPR 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.
In the new analyses, the axial alignment of a 180* rotated bundle conservatively ignores the presence of the channel fastener.
The more limiting condition of assuming that the spacer buttons are in contact with the top guide is assumed.
There is no known loading that could bend or break the channel spacer button during the insertion of a 180 rotated bundle, since bots, t'ie top guide and spacer button are chamfered to provide lead-in.
For a propec]y assembled bundle, no mechanisn exists which could invalidate the assumption that a 180 rotated bundle leans to one side.
A-1
NEDO-24567 It should be noted that proper orientation of bundles in the reactor core is readily verified by visual observation and assured by verification procedures during core loading.
Five separate visual indications of proper bundle orientation exist:
(1) The channel fastener assemblies, including the spring and guard used to maintain clearances between channels, are located at one corner of each fuel assembly adjacent to the center of the control rod.
(2)
The identification boss on the fuel assembly handle points toward the adjacent control rod.
(3) The channel spacing buttons are adjacent to the control rod passage area.
(4) The assembly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell.
(5) There is cell-to-cell replication.
Experience has demonstrated that these design features are clearly visible so loat ug that any misloaded bundle would be readily identifiable during core verification. Figures A-1, A-2 and A-3 denote a normally loaded bundle, a 180 rotated bundle, and a 90* rotated bundle, respectively. Actual experience (References A-1 and A-2) has demonscrated that the probability of a rotated bundle is low.
The new analyses procedure results show that the minimum CPR for the most limit-ing rotated bundle in the core is greater than t e safety limit.
A-2
_ REFERENCES A-1 Letter, R. E. Engel (CE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-219-77, June 1, 1977.
A-2 Letter, R. E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," MFN-457-77, November 30, 1977.
A-3
NEDO-24587 O'N/'O f*
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Rotated Bundle, 180 Degree Rotation A-5
NEDO-24587 6'
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Rotated Bundle, 90 Degree Rotation A-6
NEDO-24587 APPENDIX B Fuel Loading Error LHGR:
15.5 kW/ft B-1/B-2
NEDO-24587 Appendix C For the past several months, General Electric, with the approval of the Nuclear Regulatory Commission in cooperation with BWR Owners and EPRI, has been engaged in a program of confirmation transient testing which has resulted in the develop-ment and qualification of an improved transient model. A description of the improved transient computer model (ODYN), its qualification and its general licensing application have been transmitted to the U.S. Nuclear Regulatory Commission in References C-1 through C-4 At the staff's request, ODYN analyses of the limiting fast pressurization trans-1ents at end of cycle 4 with Recirculation Pump Trip are being supplied in this appendix.
Transients analyzed with ODYN in support of recirculation pump trip are the Load Rejection without Bypass (LR w/o BP), the Turbine Trip without Bypass (TT w/o BP), and the Feedwater Controller Failure (FWCF).
For different transients under different conditions, the ACPR calculated using ODYN may be larger or smaller than that calculated using REDY.
Table C-1 presents the results of the ODYN analysis.
The analyses presented in this appendix differ from the standard licensing calculational procedure in that the assumed initial MCPR for each transient is equal to the safety limit CPR plus the ACPR for that transient. These transient-dependent initial CPR's are given in Table C-1, and Figures C-1, C-2 and C-3 depict the transients.
C-1
Table C-1 CORE-WIDE TRANSIE?TT ANALYSIS RESULTS (ODYN ANALYSES WITH RECIRCULATION PUMP TRIP) 8x8 7x7 h
h/A SL V
Power Flow Transient
(%)
(%)
Exposure
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NEDO-24587 REFERENCES Appendix C C-1.
Letter MFN 462-77, E. D. Fuller to D. F. Ross, " Transmittal of ODYN Computer Model Description", dated December 2, 1977 C-2.
Letter MFN 058-78, E. D. Fuller to D. F. Ross, " General Electric Proposal for Licensing Basis Criteria", dated February 7, 1978 C-3 Letter MFN 014-78, E. D. Fuller to D. F. Ross, " Transmittal of Draft ODYN Qualification Report", dated January 13, 1978 C-4.
Letter MFN 136-78, E. D. Fuller to D. F. Ross, " Application Submittal for ODYN Transient Model", dated March 31, 1978 C-6
NEDO-24587 APPENDIX D
~
RECIRCULATION PUMP TRIP FEATURE D.1.
INTRODUCTION Significant 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 recircu-lation pump motors any time turbine control valve or turbine stop valve fast closure occurs. This results in a rapid reduction in recirculation flow and increases the core void content during the core-wide pressurization transients, thereby reducing the peak transient power and heat flux. A more detailed dis-cussion of the effect of RPT is included in Section D-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 recircu-lation pump motors are tripped; this releases the recirculation 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.
D.2.
_EFFECT OF RPT ON PLANT PERFORMANCE D.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 negative void reactivity coefficient permits load following through control of the recircu-lation flow without control rod movement. To increase power, core flow is increased, which decreases the void fraction and icnreases the neutron modera-tion and reactor power.
eThese are the same switches which initiate scram on control valve fast closure or stop valve closure. By using the same signal to initiate RPT, the necessary hardware modifications are minimized and the scram trip and RPT are initiated simultaneously.
D-1
NEDO-24587 The negative void reactivity characteristic of the BWR dictates the necessity The for reactivity control during certain operational pressurization events.
two most lirtiting events analyzed in a typical plant safety analysis are the rapid turbine stop valve closure (turbine trip) or control valve closure In these events, the (generator lead rejection) with assumed bypass failure.
doms pressure increases rapidly, causing a reduction in the core average void fraction, 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 reactivity to the system faster than the control rods add negative scram reactivity.
The BWR d3 sign provides a system for which reactivity changes have an inverse This void feedback relationship to the steam void content in the moderator.
eff ect 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) produces additional steam voids, which reduces the reactivity and thereby reduces the The void feedback mechanism contributes to the stable regulation of power.
core reactivity and permits load following without use of control rods by varying the recirculation flow. The practical constraints on the void coefficient are it must be large enough to prevent power oscillation due to spatial xenon that changes yet small enough that pressurization transients do not unduly lbmit plant operation.
The basic phenomenon associated with void feedback is the decrease in neutron moderation resulting from an increase in void fraction. A spectral shift in the neutron flux occurs wherein the thermal flux, and hence the fission rate, decreases and the epithermal flux, and hence the resonance capture rate, increases.
The Conversely, a decrease in void fraction causes an increase in reactivity.
void coefficient is predominantly the function of three variables for any fixed As bundle geometry:
(1) the average voids; (2) enrichment; and (3) exrosure.
each of these three parameters increases, the absolute magnitude of the void coefficient increases and becomes more negative.
D-2
NEDO-24587 For pressurization transients, the rate of flux rise is dependent on the magni-tude 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 The minimum scram reactivity occur; at end of cycle when control rods are core.
fully withdrawn from the core.
In this situation, it takes a longer time for the control roo to travel to a high importance region in the core. For this reason, the pressurization transients 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 posi-tive 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 end of cycle (EOC), the control
.>ds are withdrawn until, ideally, they are all with-drawing. Hence, the effectivencas of scram reactivity for shutdown of certain pressurization transients is decreased as the core approaches E0C conditions.
As discus 3ed 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 significantly reduced by a rapid reduction in core flow. This increases the core void 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 ti.e prompt tripping of both recirculation pumps. The RPT system described in Section D.3 has been developed to accomplish this goal.
D.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 boiling transition. This is known as the General Electric Thermal Analysis Basis (GETAB). GETAB statis-tically correlates a calculated MCPR as the condition at which less than 0.1%
of the fuel rods are expected to experience boiling transition.
This value D-3
NEDO-24587 is incorporated into the plant technical specifications as the fuel cladding integrity safety limit. An operating limit MCPR is established such that the most severe abnormal operational transient will not result in violating the safety The difference between the actual plant operating critical power ratio limit.
(CPR) and the operating limit MCPR s 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 derate 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 tarbine or generator trip without bypass is the limiting thermal event near The RP" system is intended to provide improved thermal cargin for these EOC.
limiting events.
D.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 overprotection 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 from posi-tion 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 J not initiated during this event.
In order Another GE criterion is that associated with unpiped safety valves.
to preclude steam from being blown directly into the containment, GE recommends that there be a minimum of 25 psi 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.
D-4
N EDO-24587 D.3 RECIRCULATION PUMP TRIP DESCRIPTION D.3 1 SYSTDi FUNCTION The RPT system, which is designed to improve fuel thermal margin, trips both recirculation pumps upon sensing stop valve closure or fast 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 positive 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 for the RPT system to ef.'ectively 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 i=provement will be seen from tripping of the pumps.
Analyses have been performed which demonstrate that the RPT system is made most effective by installing and tripping a line breaker between the recirculation pump drive motor / generator and the pump motor. Although a motor / generator field breaker trip has cost advantages over a line creaker, the response characteristics from such a trip do not achieve significant improvements in thermal margins.
Upon tripping the field breakers, the drive motor generator continues to momen-tarily 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.
D-5
NEDo-24587 In orden to achieve the desired improvements in thermal margins for the turbine /
generator trips, the supply current to the pump motor must be terminated in less than approximately 200 milliseconds after receipt of the signal from the switches in the turbine stop valves or in the turbine control valves. The line breaker punp trip does achieve the desired system goal.
D.3 2 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 pro-tection system functions which provide protection for the same events.
The system consists of turbine control and stop valve closure sensors, 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 occur-ing below the existing 305 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 measured variable.
The system is designed to meet the single-failure criterion such that any single trip channel (sensor and associated equipment) or system com-ponent failure shall not prevent the system from performing its intended 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 shall be clearly identified to reduce the possibility of inadvertent trip of the D-6
NEDo-24587 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 calibrating the system logic quarterly and circuit breakers once per refueling outage.
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 moto" 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 calibration during power operation will not initiate pump trip action at the system level.
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