ML19259B803
| ML19259B803 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/01/1979 |
| From: | Brugge R, Ervin A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19259B802 | List: |
| References | |
| NEDO-24179, NEDO-24179-R1, NUDOCS 7903230178 | |
| Download: ML19259B803 (44) | |
Text
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4 NED0-24179 79NED256 Class I March 1979 Revision 1 SUPPLDIINTAL RELOAD LICENS!NG SU3MITTAL
, FOR BRUNS ***ICR STEAM ELECTRIC PLAFI UNIT 2 RELOAD 2 (Recirculation Pump Trip Feature) 97, '
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Prepared: O 2 ' $ *- W A. M. Ervin. Engineer Operating Licenses II Approved. K , u#, .
R. O. Brugge, Manag,er/, (
Operating Licenses II 7903230(1f Nt.;0LI AR E NE acy ea tsECr3 Divis CN a OENE A AL E' . ECraic OOv* ANv SAN .CSE. CALIFC ANia 15':5 GENER Alh ELECTRIC
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IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY 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 Commis-sion (USNRC) for acending CP&L's operating license of the Brunswick Steam Electric Plant Unit 2. The infor=ation 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 infor=ation 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 28, 1974, and nothing contained in this document shall be construed as changing said contract. The use of this infor=ation except as defined by said contract, or for any purpose other than that for which it is intended, is not authorited; and with respect to any such unauthorited use, neither General Electric Company nor any of the contributors to this document makes any representation or war-ranty (express or i= plied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such infor=ation may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which =ay result frem such use of such in f o r=ation.
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NED0-24179-1
' 1. PLANT-UNIQUE ITEMS (1.0)*
Rota:ed Sundle Analysis Procedur.e: Appendix A Total Nu ber and, Capacity of Safe' y/ Relief Valves: Reference 2 Tuel Leading Error LHCR: Appendix 3 00YN Transien: Calcula: ion Results: Appendix C Re:irculation Pu=p Trip Fea:ure: Appendix D Separa:e MCPR Limits Repor:ed for 8 x S and 8 x SR Fuels
- 2. RELOAD FUEL BL*NDLES (1.0, 3.3.1 and 4.0)
Fuel Tvoe Nu-ber _Nu-her Drilled Irradiated Ini:ial Core Type 1 108 108 Ini:ial Core Type 3 176 176 7D3230 4 4 SDB274L 100 100 8DB274H 40 40 New .8DR3265H 64 64 SDR3283 6S 63 Total 560 560
- 3. REFERENCE CORE LOADINO PATTERN (3.3.1)
No=inal previous cycle exposure: 11,570 mwd /:
Assumed reload cycle exposure: 13,080 mwd /t Core leading pattern: Figure 1
- 4. CALCULATED CORE EFFECTIVE ML*LTI?t! CATION AND CONTRCL SYSTEM WCRTH - NO VOIOS, 200C (3.3.2.1.1 and 3.3.2.1.2)
SOC ke..,,
Uncontrolled 1.120 Fully Con: rolled 0.958 S:ronges: Control Rod Out 0.989
, R. Maxi =u= Increase in Cold Core Reactivi:y with Exposure Into Cycle, ik 0.000
- 5. STA'.~JSY LICt'!D CONTROL SYSTEM SHUTDOWN CAPA3ILIS' (3. 3.2.1. 3)
Shutdown Margin (ak) pe (20*C, Xenen Free) 600 0.032
- ( ) refs s to areas of discussion in Reference 1.
1
- 6. RELOAD UN!OCE TRA'iSII'.7 ANALYSIS INP'.75 (3.3.2.1.5 and 5.2)
EOC Void Coefficien: N/A= (c/2 Rg) 7.46/9.49 "Joid Fra:: ion (2) 41.76 Doppler 20efficien: N/A (:/* F) 0.1937/0.1 Sea Average Fuel Te:pera:ure ( F) 1538 Scras Wor:h N/A (5) 38.75/31.00 Scram Rese:ivi:y Figure 2
- 7. RELOAD UNIOUE GETA3 TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 7x7 8x3 8x8R EOC EOC EOC Peaking factors (local, radial and axial) 1.24/1.2S4/1.40 1.22/1.44S/1.40 1.20/1.585/1.40 R-Fac:or 1.100 1.098 1.051 Bundle Power 5.481 6.175 6.753 (MW:)
Sundle Flev (103 lb/hr) 124.5 113.0 114.0 Initial MCPR 1.20 1.20 1.20
- 8. SELECTED ifARGIN IMPROVEMENT OPTIONS (5.2.2)
Recirculation Pump Trip
- N = Nuclear Inpu: Data A = Used in Transien: Analysis 2
- 9. CORE
- WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Power Flov o Q/A Psi Py ACPR Plant Transient Execsure ( ?.) (%) (%) (%) (psig) (psig)_ 7x7 8x5/SxSR Reseense Turbine Trip without Sy; ass EOC3
- 104 100 133.0 100 1166 1197 0.01 0.03 Tigure 3 Inadvertent HPCI Pump 0.13 104 100 121.2 112.9 1018 1067 0.11 f, Tigure 4 Start Teedvater Centro 11er Tailure EOC3 104 100 109.3 104.8 1028 1073 0.05 0.06 Figure 5
- 10. LOCAL ROD WITEDRAWAL ERROR (WITH LIMITINO INSTRUMINT FAILURE)
Red Position Rod Block (Teet ACPR MLHGR (kW/ft) Limiting Reading '*it
. hd r a vr. ) 7x7 8xS 8x8R 7x7 8x8 8xSR Red Pattern 104 4.0 0.13 0.10 0.19 18.0 15.3 12.5 Tigure 6 105* 4.0 0.13 0.10 0.19 18.0 15.3 12.5 Tigure 6 106 4.5 0.15 0.12 0.22 18.8 16.3 13.1 Tigure 6 107 5.0 0.17 0.13 0.25 19.4 16.8 13.6 Tigure 6 108 5.5 0.20 0.14 0.27 19.8 17.3 14.1 Tigure 6 109 6.0 0.22 0.16 0.29 20.0 17.5 14.4 Tigure 6 110 9.0 0.24 0.24 0.36 18.2 16.5 14.3 Figure 6
. 11. OPERATING MCPR LIMIT (5.2)
BOC3 - ECC3 1.20 '
(8x8 fuel) 1.26 (8x8R fuel) 1.20 (7x7 fuel)
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Tlev Psi Pv Plant Transient ( P. ) (%) (psig) (psie) Response MSIV Closure (Flux Scram) 104 100 1213 1258 Tigure 7
- Indicates setpoint selected 3
- 13. STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 ,
Rasetor Core Stability:
Decay Ratio, x2 *O '0 (1052 Rod Line - Natural Circulation Power)
Channel Hydrodynanic Pt for:ance Decay Ratio, x2 /*0 (105% Rod Line - Natural Circulation Power) 8x8/8x8R channel 0.2S 7x7 channel 0.13
- 14. LOSS-OF-COOI. ANT ACCIDENT RESULTS , (5.5.2) 8DR3265 Exposure MAPLHCR PCT Local Oxidation (mwd /t) (kW/ft) ("F) 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 SDR3283 Exposure MAPLEGR PCT Local Oxidation DNd/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 12.0 219'i 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
NED0-24179-1
- 15. LOADING ERROR RESULTS* (5.5.4, Appendix A)
, Limiting Event: Rotated bundle 8DRB283H or 8DRB265H HCPR: 1.07**
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 a
- Using New Rotated Bundle Analysis Procedures described in Appendix A.
- Includes added penalty of 0.02 imposed by NRC.
5
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43 3; 6 39 32 35 3a I. 2 31 25 27 6 2 0 23 NOTES: 1. ROD PATTERN IS 1/4 CORE MIF20R SY)SETRIC.
UPPER LEFT QUADRANT SHOWN ON MAP
- 2. NL73ERS INDICATE NLMER OF NOTCHES WITHDRA*.7 OUT OF 43. 3LArs IS A WITEDR.u3 ROD
- 3. EF20R ROD IS (22,27)
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hTDO-24179-1 24 O CALeut4TED vALUE O BCUNDING VALUE FOR 280 cal /g
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NEDO-24179-1 i:
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O BousolNG VALUE FCR 280 casig COLO C BOUNOtNG VALUE 8CR 230 =al/g MSB A CALCULATED V ALUE - COLD Q CAL:ULATED V ALUE - MSB
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NEDO-24179-1 so O souno No vALUE FOR 230 w/g 80 -
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NEDO-24179-1 REFERE'CES
- 1. " General Electric Boiling 'a'ater Generic Fuel Application," NEDE-240ll-?,
Revision 3, March 1976.
- 2. Letter No. NG-77-1060 frem E. E. Utley (CP&L) to A. Schwencer (NRC),
Septe=ber 20, 1977.
e 0
O 19/20
NEDO-24179-1 APDENDIX A N E'n' BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES The bundle loading errer analyses resul:s presented in See: ion 15 in :his supple:en: are based en new analyses procedures for the rotated bundle loading error even:s. The use cf this new analysis procedure is discussed below.
A.1 NEW ANALYSIS PROCEDURI TOR THE ROTATED BUNDLE LOADING IRROR EVENT The reta:ed bundle leading errer event analyses results presen:ed in :his sup-pienen: are based on the new analyses procedure described in References A-1 and A-2. This new me: hod of performing the analyses is based on a more de: ailed analysis model, which reflec:s more accura:e analyses :han cha: used in previous analyses of this event.
The principle difference be: ween the previous analyses procedure and the new analyses precedure is the =edeling of the wa:er gap along the axial leng:h of the bundle. The previous anc;'ses used a uniform wa:er gap, whereas the new analyses utilize a variable water gap which is represen:ative of the actual tendi: ion.
The effec: of the variable wa:er gap is to reduce the power peaking and the R-factor in the upper regiens of the limiting fuel rod. This results in the cal-cula:1cn of a reduced iC?R for :he ro:ated bundle. The calculation was , performed using the same analytical models as were previously used. The cnly change is in the simula:1on of the wa:er gap, which more accurately represen:s the actual geenetry.
In the new analyses, the axial alignment of a 180' rotated bundle conservatively igneres the presence of the channel fas:ener. The more limiting condition of assumin-g that the spacer but: ens are in contact with the top guide is assumed.
There is nc known loading that could bend or break the channel spacer button during the insertion of a 180* rotated bundle, since both the top guide anJ spacer bu::or. are chamfered :o provide lesd-in. For a properly assembled bundle, no mechanist- exists which could invalidate the assumption ; hat a 180* rotated
, bundle leans to one side.
A-1
NEDO-24179-1 1: should be noted :ha: proper orien:':icn a of bundles in :he rea::or core is readily verified by visual observa:1cn and assured by verifica: ion procedures during c re leading. Five separate visual indica: ions of proper bundle or:en:a:i:n exis::
(1) The channel fas:ener assemblies, including the spring and guard used to =ain:ain clearances be:veen chanrels, are located at one corner cf each fuel assembly adjacen: to the can:er of the control red.
(2) The iden:ification boss on the 'ael asse:bly handle points :ovard :he adjacen: con:rel rod.
(3) The chanrel spacing bu:: ens are adja:en: :: the con:rel red passage area.
(4) The assembly iden:ifica:ica numbers which are loca:ed en :he fuel asse=bly handles are all readable from the dire::icn of :he cen:er of the cell.
(3) There is cell-to-cel: replica:1en.
Experita:e has de=enstraced tha: these design fea:ures are clearly visible se cha: any =isleaded bundle would be readily iden:ifiable during core leading verification. Figures A-1, A-2 and A-3 denc:e a normally loaded bundle, a 15;*
re:a:ed bundle, and a 90* rota:ed bundle, respec:ively. Actual experien:e (References A-1 and A-2) has de=enstra:ed tha: the probability of a rota:ed bundle is lov.
The new analyses procedure results show tha: the mini =um CPR for the =cs: 11=1:-
ing rotated bundle in the core is grea:er than :he safe:y 11=i:.
A-2
NEDO-24179-1 Rer. r_RrN Cr__e
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- o D. Eisenhu: (!;RC) , " Fuel Assembi-; Loading
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A-2 Let -:, R. E. Engel (CE) te D. Eisenhut (NRC) , "Fu e l As s e"~x I v -
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O,'lO NOTE BUNDLE NUM8ERS ARE FOR ILLUSTRATsvE PURPOSES ONLY Figure A-1. Nor=al Loading A-4
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NEDO-24179-1 o\
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o[lo NOTE. SUNOLE NUMBERS ARE FOR BLLUSTRATIVE PURPCF;5 ONLY Tigure A-3. Rotated Sundle, 90 Degree Rotation A-6
4 NED0-24179-1 APPENDIX B Fuel Loading Error LHGR: 15.5 kW/ft B-1/B-2
NEDO-24179-1 APPENDIX C Fer the pas: several men:hs, General Electric, vi:h the approval of the Nuclear Regula:rrv C:= issi:n in coopera:icn vi h 3k'R Owners and IFR1, has been engaged in a prepra: cf confir a:1cn : ansien: testing uhich has resulted in the develc;-
=en: and qualift:a:ica ef an improved ::ansient codel. A des:ription of the i=preved transien: computer model (CDYN), its qualifica:icn and its Feneral licensing application have been trans=i::ed to the U.S. Nuclear Regula:ory Cc--issien in References C-1 through C 4 A: the staff's reques:, ODYN analyses of the li:1:ing fast pressuriza:icn trans-ients at end of cycle 4 with Retirculation Pump Trip are being supplied in :his appendix. Transien:s analyzed with ODYN in supper: of recirculatica pump tri; are the Lead Rejec:ic: withou: Sypass (LR v/c 3?), the Turbine Trip vi:hou:
Bypass (TT v/o 3?), and the Feeduater Controller Failure ( F.'CT) . For differen:
transien:s under different conditions, the ACPR calcula:ed using ODYN =ay be larger er stalier chan tha: calcula:ed using REDY. Table C-1 presents the results of the ODYN analysis. The analyses presented in this appendix differ f c= :he s:andard licensing calcula:ienal procedure in tha: the assumed ini:ial
..CPR w for each :ransien: is equsi to :he ssfaty liri: CPR plus the aCFR fer tha:
transient. These transient-dependent initial CPR's are given in Table C-1, and Figures C-1 and C-2 depict the transients. .
C-1
NEDO-24179-1 Table C-1 CORE-WIDE TRANSIENT ANALYSIS RESULTS (ODYN ANALYSES WITH RECIRCULATTON PlRtP TRIP)
Power Flow $ QA P3t P'y 8x8/8x8R 7x7 Transient exposure (%) (%) (% initial) (% initial) (psia) (psig) ACPR ACPR Turbine Trip EOC3 104 100 280.6 107.2 1187 1212 0.12 0.08 without Bypass Feedwater EOC3 104 100 10E.3 106.0 1070 LJutroller 1093 0.07 0.05 Fa ilure C-2
I I I NEUTRON FLUX l 1 VESSEL Pf ES ntSE IPSil 1 2 RVE SURffCE HEAT FLUX 2 SAFETY VI LVE FLOW 150' 3 CORE INLE T FLOH 300* 3 f1EL IEF VI (JE FLOW _
4 4 BIPRSDI LVE7 LOW 5 S G
g100. K 200.
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TIME ISECl TIME (SECl 5
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I LEVELilNOf-REF-SEP-SKIRT I VOID REfV TIYlTT ae 2 VESSEL S1EANFLOW /\ 2 (MPPLEA I EnCTIVITT H 200' 3 TURBINE f TERMFLOH y* / \ ~ _ 3 SCRAM REI CTIVliY TTEEOMTEF FLDH 1I 10IT~llfiCllVliF t
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TIME (SECl IIME (SECl Figure C-1. Plant Response to Turbine Trip Without Bypass, Trip Scram (ODYN Analysis with RPT)
i 14 tillith 6 I 111 1 VI$SIL s1 i S Hl'J H*Sil 2stvl Msue i1 it III iLUX 2 NilI I T VI i vt ttIM 193, INI l 1 ) Id l lQ , VI I Vi l10!.
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liti IMil Ilti 15101 Figure C-2. Plant Response to Feedwater Controller Failure
NEDO-24179-1 REFERENCES C-1. Letter MFN 462-77, E. D. Fuller to D. F. Ross, "Trans=ittal of ODYN Cenputer Model Description," dated Dece=ber 2, 1977 C-2. Letter MFN 058-78, E. D. Ful~ to D. F. Ross, " General Electric Proposal for Licensing Basis .riteria," 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 Sub=ittal for ODYN Transient Model," dated March 31, 1978 c-5/c-6
NEDO-24179-1 APPENDII D RECIRCULATION PUMP TRIP FEATJP.E D.1 INTRODUCTION Significant i= prove =ent in ther=al cargin can be realized if the severity of core-wide pressuri:ation transients is reduced. The Recirculation Pu=p Trip (RPT) feature acco=plishes this by rapidly cutting off power to the recircu-lation pu=p :oters any time turbine control valve or turbine stop valve fast closure occurs. This results in a rapid reduction in tecirculation flow and increases the core void content during the core-wide pressurization transients, thereby reducing the peak transient power and heat flux. A =cre detailed dis-cussion of the effect of RPT is included in Section D-2.
Basically, the RPT syste: 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 =otor generator sets and the recircu-la_tien pu=p meters are tripped; this releases the recirculation pu=ps to ceast down under their inertia. Adding RPT will result in a reduction in CPR for transients involving stop valve or centrol valve closures.
D.2 EFFECT OF RPT ON PL'.NT 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 =ederation, which is represented by a negative void reactivity coefficient. This negative void reactivity coefficient per=its load following through control of the recircu-lation flow without control rod =ove=ent. To increase power, core flow is increased, which decreases the void fraction and ienreases the neutron =edera-tion and reactor power.
'These are the sa e switches which initiate scra= on control valve fast clesure er stop valve closure. By using the sa=e signal to initiate RPT, the necessary hardware =edifications are =ini=ized and the scra: trip and RPT are initiated si=ultaneously.
D-1
NEDO-24179-1 he negative voic reactivity e.aracteristic cf the SWR dictates the necessity fer reactivity centr:1 during certain' operatienal pressuri:ation events. Se tvc =cs: 11=iting events analy:ed in a typical plant safety analysis are the ra;it tuttine step valve elesure (turtine trip) er cont ci valve closure (genera:Or lead ejectien) with assu=ed i/ pass failure. In these events , the d:ce pressure increases rapidly, causing a reduction in the cere average void fraction, which increases =ederation and results in a pcsitive power increase.
This is reflected in decreased =argins to pressure and themal 11=1:s.
The physical phenomenon which causes the reduced =argins is that the void reactivity feedback, which is due to the pressurization, =cmentarily can sdd positive reactivity to the system faster than the centrol rods add negative scra= reactivity.
"he 3WR design prevides a systes for which n. activity changes have an inverse relationship to the steam void centent in the moderator. "his void feedback effect is cne of the inherent safety features of the 3WR syste=. Any systes input which increases reacter pcwer (either in a local cr gress sense) produces additional steas voids, which reduces the reactivity and thereby reduces the power. "he void feedback mechanis= contributes to the stable regulation of core reactivity and per=its load folicwing without use of centrol rods by varying the recirculatica flew. The practical constraints en the void coefficient are that it =ust be large e.cugh to prevent pcwer escillation due to spatial xenen changes yet s=all enough that pressurization transients do not unduly li=it piant operatien.
The basic phenomenc associated with void feedback is the decrease in neutron moderation resulting frem an increase in void fraction. A spectral shift in the neutren flux occurs wherein the themal flux, and hence the fission rate, decreases and the epither=al flux, and hence the resonance capture rate, increases.
Conversely, a decrease in void fraction causes an increase in reactivity. The void coefficient is predccinantly the function cf three variables ter any fixed bundle seemetry: (1) the average voids; (2) er-ichmen t; ant (3) expcsure. As each of these three parameters increases, the absolute =agnitude of the void coefficient increases and beccces =cre negative.
D-2
NEDO-24179-1 For pressurizatien transients, the rate of flux rise is dependent en the =agni-tude of the void coefficient. The = ore negative the void coefficient, the greater the flux rise rate. The rate at e.ich the negative reactivity can be added to the ecre by the scra: deter =ines the severity of the transient. The sera: reactivity depends en the ability of the control rods to be in the high flux regions of the core. The =inimum sera: reactivity occurs at end of cycle when control rods are fully withdrawn fro: the core. In this situation, it takes a lenger time for the centrol rod to travel to a high i=portance region in the core. For this reason, the pressuri:ation transients are : st severe near thc end of the cycle.
The degree to which the pressure and the.~::al sa gins are reduced during pres-suri:ation events _ depends en the tradeoff between the negative scra: and posi-tive void reactivities. Typically, at beginning of cycle (BOC), control rods are partially inserted; this per=its a pro =pt shutdown of the syste: without a significant decrease in =argins. As the fuel cycle proceeds toward end of cycle (E00), the control rods are withdrawn until, ideally, they are all with-drawing. Hence, the effectiveness of scra: reactivt3y for shutdown of certain pressurization transients is decreased as the core approaches ECC conditions.
As discussed above, =argins are decreased when the positive void reactivity feedback is inserted at a rate faster than the negative scra: reactivity.
Analyses have shown that the transient severity can be significantly reduced by a rapid reductien in core flow. This increases the core veid fraction' during pressurization transients and consequently =inteises the power rise experienced.
The rapid reduction in core flow necessary to acco:plish this effect can be achieve: cy the prc=pt tripping of both recirculation pu ps. The RPT syste:
described in Section D.3 has been developed to accc plish this goal.
D.2.2 Thermal Limits Consideration One of the operating fuel ther:al limits, the =ini=um critical power ratio (M0?R),
is established such that the mest severe abnor:al operatienal transient is not expected to subject = ore than 0.1% of the fuel rods to boiling transition. This is known as the General Electric Ther:al Analysis Basis (GETA3). GETA3 statis-tically correlates a eniculated M0?R as the cendition at which less than 0.1%
of the fuel rods are expected to experienee boiling transition. This value D-3
NEDO-2/.179-1 .
is incorporated int: the plant technical specifiesticas as the .^.:e1 claddi..g integrity safety limit. An operating li=it MC?R is established such that the sost severe abn - 21 ope-ational transient vill not result in viciating the aafety
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.w.. .. . a *a. .ua .' a '. a**
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. ..**.*..*2'. pcve. .a*..*
(C?R' a..d the operating limit M"?R s a sensu. e of the the. - ~
al =a. gin.
If the nor=al ope-ating CPR at the licensed power level ca . net be =aintained above the ope-sting limit MC?R, a plant derate vill te i=; sed to assure that the resultant change in CPR frc: a worst-case abnor-.a1 ope. ational t. ansient -
vill not decrease the MC?R bel:w the safety li=1t. A reduction in severity of the verst transient allows a reducticn in the cpe-ating 11:10. Usually either a turbine or generator trip without bypass is the 11:1 ting the al event nea.-
E00. The R77 syste: is intended to provide i=;-eved the. a1 =argin fc.- these -
...,.. . eve..s.
D.2.3 Overpressure Prctection Considerations The R?! sys:e has no effe:: en overpressure prete :icn considerations since it i.i not initiated during :he even: (MS!'.' Closure vich Indire:: Scra=) which de= ens:: aced ::=;11ance -1:5 the ash;. vessel everpressure prete :icn 11=1:.
D.3 RECIRCULATION PUMP TRIP DESCRIP H ON D.3.1 System Function
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re:irculation pu=;s upc= sensing step valve closure er fast centrol valve closure. he reduced core flow reduces the void ec11 apse in the ecre during two cf the :st li=iting pressurination events (i.e., turbine and ger.erator trips). T.-ipping cf the reci-culation pu=;s results in a s= aller net ;csitive void reactivity addition to the syste: during these pressurizatien events.
Tc.is results in a lower pcwer increase and consequently a lower Operating M ??
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decrease in ther:21 =argins, the effect of reduced flew en the power increase is a C0. side.atly : re d::inant effect and the ne* result is to Ptduce the the.~ al severity of the eve.t.
DJ
NED0-24179-1 In ceder for the R?! syste: to effectively counteract the void collapse effects fre= , pressurization transients, the pc:p trip =ust occur very soon after the turbine / generator trip, and the pu=ps must coast down at a relatively fast rate.
If the p==p trip and coastdown do not occur quickly, the positive void reactivity feedback caused by the pressurization effects will de=inate the transient and no =argin i= prove:ent will be seen frc: tripping of the pu=ps.
Analyses have been perfor:ed which demonstrate that the RPT system is made mest effective by installing and tripping a line breaker between the recirculation pu=p drive motor / generator and the pu=p motor. Although a motor / generator field breaker trip has cost advantages over a line breaker, the response characteristics fro: such a trip do not achieve significant improve =ents in ther=al =argins.
Upon tripping the field breakers, the drive motor generator continues to me=en-tarily supply some reduced power to the pu=p motor due to the time required for the generator field and line current to drop to zero. This res;lts in reduced effectiveness of the system.
In order to achieve the desired i=; rove =ents in ther=al cargins for the turbine /
generator trips, the supply current to the pu=p motor must be ter=inated in less than approxi=ately 200 milliseconds af ter receipt of the signal frc: the switches in the turbine step valves or in the turbine control valves. The line breaker pump trip does achieve the desired syste= goal.
D.3.2 System Description The RPT system includes all equipment that trips recirculation pa=p motors fro:
their power supplies in response to a turbine / generator trip or load rejection.
The RPT syste: is designed to be of quality consistent with the reactor pro-tection syste= functions which provide protection for the sa:e events. The syste: consists of turbine centrol and stop valve closure sensors, separate division logie and two circuit breakers for each pu=p motor. The RPT syste: is designed to be operable whenever the turbine generator trip scra: is operable (i.e., abeve approxi=ately 30% reacter ther:a1 pressure). Exis ting turbine first-stage pressure senscrs will prevent R?T initiation fcr turbine-generat:r trips occurring below the existing 305 power bypass of turbine and generat:r trip sera: si gnals .
D-5
NEDO-24179-1 he RPT soste= design includes two separate trip divisiens with each having twc separate trip channels, sensors and associated equip =ent for each =easured va-iatie. The syste= is designed to =ett the single-failure criterion such that any single trip channel (sensor and associated equip =ent) or syste: ec=-
ponent failure shall not prevent the syste= frc= perfor=ing its intended safety function. Electrc=echanical relays used as the logic ele =ents within the syste=
and the syste= logic are of the failsafe type (i.e., trip en Icss of electrical power).
The RPT system is designed to acco=plish the desired function and to =inimize the effect of this additional syste= cn plant availability. The syste= logic is designed such that it will not cause the inadvertent trip of = ore that. ene pu=p given a single co=penent failure in the syste=. Each trip divisien shall be clearly identified to reduce the possibility of inadvertent trip of the recirculation pu=p during routine =aintenance and test operations. Redundant sensor circuits in each division (sensors, viring, trans=1t 4r, a=plifiers, etc.) are electrically, =echanically, and physically independent so ttat they are unlikely to be disabled by a ecc=en cause except for an electrical pcwer failure.
Capability is provided for testing and calibrating the syste= logie quarterly and circuit breakers once per refueling outase. Provisions are =ade to. allev closure of stop valve and fast closure of turbine entrol valve separately at lesst ene valve at a ti=e (for no:..al routine valve test purposes) without causing a pu=p motor trip. *he syste= input sensors and the divisien logic are capable of being checked one channel er divisicn at a ti=e. The sensces and syste: logic test or calibration during power operation will not initiate pu=p trip action at the system level.
D.4 RPT LOGIC DIAGRAMS, CIRCUITS AND TESTABILITY m e, .,,ose of RPT is to reduce the severity of the reactivity transient caused by either of two postulated events: (1) turbine trip with failure of the bypass valve, and (2) generator load rejection with failure of the bypass valves.
D-6
NED0-24179-1 RPT is not required for any other postulated events; therefore, the logic begins with the sensing of stop valve closure (turbine trip) or fast closure of the turbine control valves (generator load rej ection) . RPT is initiated following either event for turbine power levels greater than 30%, but is independent of the operation of the bypass valves.
Figure D-1 is the logic diagram for RPT System A that trips pu=p A. System B is the same except for the nomenclature changes indicated in parentheses.
The logic is "tuc-out-of-two or two-out-of-two" for both stop valve closure and s fast closure of the turbine control valves. This means, for example, that the closure of stop valves 1 and 2 will trip both recirc pumps through System A or stop valves 3 and 4 will trip both pu=ps through System B. This logic provides the testability feature with which any one stop valve or control valve can be closed and system status tested by observing relay contact status without causing RPT operation. The entire logic of one division of the RPT system may be tested without tripping the pumps by placing that system control switch briefly in the "inop" position for the duration of the test; the test is initiated by closing the two stop valves which initiate that system to the 10% closed (90% open) position. Successful completion of the test is indicated by annunciation of "RPT initiate" as the annunciation relays are energized. During this brief inter-val, the redundant RPT system is, of course, centinuously available to perform its safety function.
k)
D-7
NEDO-24179-1 TURSINE CONTROL TURBINE STOP VALVE A (C) VALVE 1 (3) <90%
FAST CLOSURE OPEN AUX DEV CR AUX DEV CR 1r IP PERMISSIVE WH EN TURBINE CONTROL PERMISSIVE WHEN VALVE 8 (D) FAST TURSINE STOP VALVE CLDEURE PRESENT 2 H) <M MN CR AUX DEV
\ AUX DEV CR AUX DEV- AUX 8UARY DEV:CE CR - CONTROL ROOM
=
RMS-REMOTE MANUAL SWITCH if PS- PRESSURE SWITCH PERMISSIVE WHEN INITIATING OR PhlAL TURRINE 1 ST STAGE ACTION PMESS 230% OF RATED POWER f x
) PERMIS$1VE ACTION A (C)
LOCAL [
1P PERMISSIVE WHEN TURBINE 1 ST STAGE BREAKER CONTROL BRE AKER CONTROL SWITCH IN CLOSE SWITCH IN CLCSE M ESS M OF POSITION RATED POWER POstTION LOCAL
\ B (D) \ RMS CR
[ \ RMS CR
[
1r PERMISSIVE WHEN
- REACTOR PUMP TRIP SWITCH IN NORMAL POSITION
\ RMS LOCAL [
p 1r 1r y TRIP Coll OPEN CLOSING CCell TRIP Coll OPEN Cl,OSING COIL CIRCUIT BREAKER C83A IC84A) CIRCUlT BREAKER C338 (CS48)
Figure D-1. Recire Pump Trip System A Typical for System B & cept as Shown ( )
D-8