ML20212R231
| ML20212R231 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 01/23/1987 |
| From: | ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | |
| Shared Package | |
| ML17303A252 | List: |
| References | |
| NUDOCS 8702020558 | |
| Download: ML20212R231 (128) | |
Text
{{#Wiki_filter:INDEX C1 DEFINITIONS - SECTION PAGE
- 1. 0 DEFINITIONS 1.1
- 1. 2 ACTI0N.............:.........................................
1-1 AX IA L S HAP E I ND EX.r. . . . . . . . . . . . . . . . . . . . . . . . . . .1-1 .............
'1.3 AZIMUTHAL POWERq TILT - T .................................... 1-1 1.4 CHANNEL
- 1. 5 CALIBRATION......................................... 1-1 CHANNEL
- 1. 6 CHECK............................................... 1-1 CHANNEL FUNCTIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . 1-2 .............
- 1. 7 CONTAINMENT 1.8 INTEGRITY....................................... 1-2 CONTROLLED LEAKAGE............. ............................ 1-2
. 1. 9 CORE 1.10 ALTERATION............................................. 1-2 DOSE EQUIVALENT I-131...............................,.i.....;
1-3 1.11 E - AVERAGE DISINTEGRATION , ENERGY........................... 1-3 1.12 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.................... 1-3 F R E Q U E N CY N0T AT I O N . . . . . . . . . . . . . . . . . . . . . . . . 1-3 ............... 1.14 GASEOUS RA0 WASTE SYSTEM..................................... 1-3 1 15 IDENTIFIED f.g g# 4 LEAKAGE.......................................... 1-3
> /-4 1.1Jr7 MEMBER (S ) 0F THE PUBLIC. . . . . . . . . . . . . . . . . . . . . . . . 1-4 .............
1.1/00FFSITEDOSECALCULATIONMANUAL(0CDM)...................... 1-4
- 1. ]g.' OPERAB LE - OPERAB I LITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 ......
1.J93:0PERATIONAL MODE - M00E..................................... 1-4
- 1. R0 / PHYS I C S T E S TS . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 1.2f2 PLANAR RADIAL PEAKING FACTOR - F ......................../. '
1-4 - 1 423 PRESSURE BOUNDARY LEAKAGE................................... 1,4'5'
- 1. 2f4 PROCESS CONTROL PROGRAM (PCP). . . . . . . . . . . . . . . . . 1-5 .............,
- 1. 24'(P U R G E - P U R G I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5
- 1. 25'6 RATED THERMA L P0WE R . . . . . . . . . . . . . . . . . . . . . . . . 1-5 ...............
8
- 1. 267 REACTOR TRIP SYSTEM RESPONSE TIME. . . . . . . . . . . . . .1-5 ............
M' 00 1.27'd REPORTABLE EVENT...................................'.......,.. 1-5 S8 1.2B? SHUTDOWN MARGIN.............................................' 1-//c, et e 1. 2 %'$5 I T E B O UND A RY . . . . . . . . . . . . . . . . . . . . 1-6 ............. 1.30'l o S0FTWARE............................................ . . . . . . , 1-6 . N UW NS PALOVERDE-UNIT /l I . t.. ..
INDEX DEFINITIONS
...- (
SECTION PAGE 1.3f:tSOLIDIFICATION.............................................. 1-6 1.3f3SOURCECHECK................................................1-6
- 1. 3I4 STAGGER ED TEST 3 ASIS . . . . . . . . . . . . . . 1-6 .........
1.3/fTHERMALP0WER............................................... 1-6
- 1. 3g6 UNIDENTIFIED LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-# . . .~7. . . . . . .
1 46'7 UNRESTRICTED AREA. . . . . . . . . . . . . . . . . . . . . . .1-6-7 .......... 1.378VENTILATIONEXHAUSTTREATMENTSYSTEM................... 1-7 1.3,8'yVENTING.................................................... 1-7 ( e d PALOVERDE-UNITfI II.
- 1. * ' '
~ INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREME (
SECTION PAGE 3/4.0 APPLICABILITY.............................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL j . g,,g y gg7gp 210 F SHUTDOWN MARGIN Q. - Tccid
},d .>m'. w/77/a94ws/ . . . . . . . . .3/4 . . 1-1 Cold 1210*F ....................... 3/4 1-f z SHUTOOWN MARGIN -T MODERATOR TEMPERATURE COEFFICIENT..................... 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY...................3/4 1-6 3/4.1.2 B0 RATION SYSTEMS FLOW PATHS -
SHUTD0WN......................... 3/4 1-7 F LOW PATHS - 0 P ERAT ING . . . . . . . . . . . . . . . . . . . . . . . ,...... CHARGING PUMPS - 3/4 1-8 CHARGING PUMPS - SHUTD0WN............................. 3/4 1-9 .. 0PERATING......................... .. 3/4 1-10 BORATED WATER SOURCES - SHUT 00WN...................... 3/4 1-11 / B0 RATED WATER SOURCES - OPERATING............. ....... BORON DILUTION ALARMS......................... 3/4 1-13 3/4 1-14 ( 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA P0SITION.......................................... 3/4 1-21 POSITION INDICATOR CHANNELS - OPERATING...............3/4 1-25 POSITION INDICATOR CHANNELS - SHUTDOWN. .............. 3/4 1-26 CEA OROP TIME.......................... ............ 3/4 1-27 SHUTDOWN CEA INSERTION LIMIT.......................... .. 3/4 1-28 REGULATING CEA INSERTION LIMITS....................... 3/4 1-29 e PALOVERDE-UNIT /l IV
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLAN _SECTION PAGE ELECTRICAL POWER SYSTEMS (Continued) 3/4.8.2 D.C. SOURCES OPERATING.~............... SHUTD0WM? ............... ........................... 3/4 8-9 3/4.8.3 ........................... 3/4 8-13 ONSITE POWER DISTRIBUTION SYSTEMS . OPERATING............. SHUTD0WN.............. .............................. 3/4 8-14 3/4.8.4 .............................. 3/4 8-16 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PROTECTIVE PENETRATION CONDUCTOR OVERCURRENT DEVICES................................. 3/4 8-17 MOTOR-OPERATED AND BYPASS VALVES THERMAL OVERLOAD PROTECTION DEVICES................................. 3/4 8-40 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON _ 3/4.9.2 CONCENTRATION..................................... 3/4 9-1 INSTRUMENTATION...................................... C 3/4.9.3 DECAY TIME...........................................
.. 3/4 9-2 3/4 9-3 3/4.9.4 ..
3/4.9.S CONTAINMENT BUILDING PENETRATIONS....................... 3/4 9-4 3/4.9.6 COMMUNICATIONS.......................................... 3/4 9-5 REFUELING 3/4.9.7 MACHINE....................................... 3/4 9-6 3/4.9.8 CRANE TRAVEL - SPENT FUFL STORAGE POOL 3/4 BUILDING. 9-7
*SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LOW WATER LEVEL..................................... 3/4 9-8 3/4.9.9 LEVEL...................................... 3/4 9-9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM................ 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES..
CEAs............. ................................... 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE
................................... 3/4 9-12 3/4.9.12 P00L.............................. 3/4 9-13 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM.............. 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS ,
3/4.10.1 SHUTDOWN MARGIN. 4.tN. ..... . . . ./. . .CcW &. . .#7#' ' ,,,
............... W57.s3,4 19_7 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION 3/4.10.3 REACTOR COOLANT LIMITS............. 3/4 10-2 L00PS................................... 3/4 1,0-3 PALOVERDE-UNIT /i IX-
INDEX _ LIMITING CONDITIONS FOR OPERATION AND SURVEILLAN
&SECTION 10,1$ (J/VT1df}L GlPLUll1 DOL! TGfl% PM2j3V1
& 10 -10 PAGE 3/4.10.4 f CEA POSITION, REGULATING CEA INSERTION LIMITS
{ 3/4.10.5 AND REACTOR COOLANT COLD LEG TEMPERATURE. 3/4 10-4 .... 3/4.10.6 MINIMUM TEMP.ERATURE AND PRESSURE FOR CRITICAL 3/4 10-5 SAFETY INJECTION 3/4.10.7 TANKS...................,............... 3/4 10-6 SPENT FUEL POO L LEVEL. . . . . . . . . . . . . . . . . .3/4 3/4.10.j SAFETY . . 10-7 3 .to., MV7%WA> INJECTION TAN M4)?Gw.K 4 PRESSURE. 0 K,ci.,- ca .z. w
. .3
. . 745
. . . . pug
. . . . . . . . . .3/4
. . .10-8 3 .11 RADIOACTIVE EFFLUENTS ,,, ,_ 9 3/4.11.1 SECONDARY EVAPORATION PONDSSYSTEM LIQUID WASTE DISCHARGES TO ONSITE CONCENTRATION................................ .......... 3/4 11-1 00SE......................................... .......... 3/4 11-5 LIQUID HOLDUP TANKS..................................~... 3/4 11-6 ~
3/4.11.2 GASE0US EFFLUENTS DOSE RATE............................................... 3/4 11-7 ( DOS E - NO B LE GAS ES. . . . . . . . . . . . . . . . . . . . . 3/4 . . .11-11 DOSE - 10 DINE-131,10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE ~ FORM..................... 3/4 11-12 GASEOUS RADWASTE TREATMENT.............................. 3/4 11-13 EXPLOSIVE GAS MIXTURE................................... 3/4 11-14 GAS STORAGE TANKS....................................... 3/4 11-15 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................. 3/4 11-16 3/4.11.4 TOTAL D0SE.............................................. 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... 3/4 12-1 3/4.12.2 LAND USE 3/4.12.3 CENSUS......................................... 3/4 12-11 INTERLABORATORY COMPARISON PR0 GRAM...................... 3/4 12-12 1 PALO VERDE - UNIT [l X . l
.1
INDEX BASES
' l SECTION-3/4.9.6 REFUELING
_PAGE MACHINE....................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P0OL BUILDING. B 3/4 9-2 . 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION........... B 3/4 9-2 3/4.9.9 CONTAINMENTTURGEVALVEISOLATIONSYSTEM............... B 3/4 9-3 3/4.9.10 and 3/4.9.11 WATER LEVEL - STORAGE POOL . . . . ..-...........................
. . . . . . . . . REACTOR VESSEL and B 3/4 9-3 3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM.............. B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS su;O x 3/4.10.1 SHUTDOWN MARGIN A . . . . . .y , -cc4 Svw?W ~75~.SB
' ................................. B 3/4 10-1 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS................ B 3/4 10-1 3/4.10.3 REACTOR COOLANT L00PS................................... B 3/4 10-1 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE. B 3/4 10-1 . . . . . .[\ . 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY..... B 3/4 10-1 3/4.10.6 SAFETY INJECTION TANKS.................................. B 3/4 10-2 3/4.10.7 SPENT FUEL POOL LEVEL................................... B 3/4 10-2 3/4.10.8 SAFETY INJECTION TANK PRESSURE.......................... B 3/4 10-2 34.10.9 sicr.ctwo MfRGW 4w 4-,- ccons -raSiw& 3/4.11 RADI0 ACTIVE EFFLUENTS g g ,o _2 3/4.11.1 SECONDARY SYSTEM LIQUID WASTE DISCHA EVAPORATION P0NDS...................RGES .................... TO ONSITE B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................. B 3/4 11-5 3/4.11.4 TO TA L 0 0 S E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B. .3/4
. . 11-6 s
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-2 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 PALOVERDE-UNIT {t XIV-
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4 lol6 K N1 ~___~_.-x._--- m DEFINITION bfinserted' ully withdrawn. full-length control element assembly of p%Wt(H)Y
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p g #{ MEMBER (ST OF THE PUBLIC l i N 1.pf7 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. employees this categorof the licensee, its contractors, or vendors.This category does not inclu Also excluded from deliveries. y are persons who enter the site to service equipment or to make for recreational, plant. 4 occupational, or other purposes not asso _OFFSITE DOSE CALCULATION MANUAL (ODCM) ' 1.1'/6TheOFFSITEDOSECALCULATIONMANUALshallconta and parameters used in the calculation of offsite doses due to radioactive
' monitoring alarm / trip setpointsgaseous and liquid effluents, in the calcula i radiological monitoring program., and in the conduct of the environmental OPERABLE - OPERABILITY 1.18'9A system, subsystem, train, component, or device shall be OPER have OPERABILITY when it.is capable of performing its specified function and when all necessary attendant instrumentation, controls, electrical power, -
cooling or seal water, lubrication or other auxiliary equipment that are
- required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE - MODE (r I 1.192 tan OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclu co'mbination of core reactivity condition, power level, and cold leg reactor coolant temperature specified in Table 1.2. PHYSICS TESTS 1.2,0'/characteristic:: nuclear PHYSICS TESTS shall be those tests performed to measu of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized un~ der the provisi ' of 10 CFR 50.59, or (3) otherwise approved by the Commission. PLANAR RADIAL PEAKING FACTOR - F xy 1.21'OThe PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to pla excluding the effects of azimuthal tilt. average power density of the individual s.~ PRESSURE BOUNDARY LEAKAGE 1.22'3 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam leakage) body, pipethrough wall, oravessel nonisolable wall. fault in a Reactor Coolant System component PALOVERDE-UNITf( 1-4. 1 -
- _ _ _ __ _a
-_. - - -. . _ - -__ - _ _ = .- .
DEFINITIONS _ PROCESS CONTROL PROGRAM (PCP) ~ 1.23'4 The PROC $55 CONTROL PROGRAM shall contain the provisions the SOLIDIFICATION of wet radioactive wastes results in a was properties radioactivethat waste meet the requirements disposal sites. of 10 CFR Part 61 and of low level t influencing SOLIDIFICATION such as pH, oil contentThe PCP shall identify pro 2 ratio of solidification. agent to waste and/or neces,sary additives for eachH O type of anticipated waste, and the acceptable boundary conditions for the process scale andparameters full scale testing shall be identified for each waste type, based on laboratory or experience. The PCP shall also include an identification of conditions that must be satisfied, based on full-scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within i the sites. limits of 10 CFR Part 61 and of low level radioactive waste disposal PURGE - PURGING 1 1.2,4'fPURGE or PURGING shall be the controlled process of discharging air or
- gas from a confinement to maintain temperature, pressure, humidity, concentra-tion, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
RATED THERMAL POWER A l 1.2/6. the RATED reactor coolant of THERMAL 3800 MWt. POWER shall be a total reactor core hea REACTOR TRIP SYSTEM RESPONSE TIME 7 1.2d The REACTOR TRIP SYSTEM RESPONSE TIME shall when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism. REPORTABLE EVENT 1.2[6 A REPORTABLE EVENT shall be any of those conditions specified in Sections 50.72 and 50.73 to 10 CFR Part 50. SHUTDOWN MARGIN. 1.2Jf/ SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: a. No change in part-length control element assembly position, and b. All full-length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn. PALO' VERDE-UNITf\ s 1-5 __.-____-_')
1 l l DEFINITIONS i SITE BOUNDARY 1.h]5fhe SITE BOUNDARY shall be that line beyond which t owned, nor leased, nor otherwise controlled by the licensee. SOFTWARE 1.3// The digital computer SOFTWARE for the reactor protection sys the program codes including their associated data, documentation, and pr . SOLIDIFICATION 1.38SOLIDIFICATIONshallbetheconversionof systems to a homogeneous (uniformly distributed), monolithic, immobilizeda outline on all sides (free-standing). solid with definite volume and SOURCE CHECK 1.3/3 A SOURCE CHECK shall be the qualitative assessment of chann when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.3)4 A STAGGERED TEST BASIS shall consist of: a. A test schedule for n systems, subsystems, trains, or other designated into n equalcomponents subintervals,obtained and by dividing the specified test interval b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval. THERMAL POWER 1.3/5 coolant. reactor THERMAL POWER shall be the total reactor core heat tr
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UNIDENTIFIED LEAKAGE ._ 1.3/6 UNIDENTIFIED LEAKAGE shall be all leakage which does not con either IDENTIFIED LEAKAGE or reactor coolant pump controlled bleed-off flow. . UNRESTRICTED AREA . 1.36'?An UNRESTRICTED AREA shall be any area at or beyond the SITE access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any commercial, institutional, and/or recreational purposes. area w PALOVERDE-UNITfl . 1-6
OEFINITIONS VENTILATION EXHAUST TREAlt4ENT SYSTEM 1.3/8AVENTILATIONEXHAUSTTREATMENTSYSTEMshallbeanysy installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not cEnsidered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considere to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING 1.367 VENTING shall be the controlled process of discharging air or gas confinement to maintain temperature, pressure, humidity, concentration, or other operating provided condition, or required in such a manner that replacement air or gas is not during VENTING. imply a VENTING process. Vent, used in system names, does not (
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PALOVERDE-UNITfl l . t. - -
- TABLE 2.2-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS m k _ FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 7
- 2. Logarithmic Power Level - High (1) q a. Startup and Operating
- 0. o tot, 0. 011
, < Or798% of RATED <-Gr895 of RATED
- THERMAL POWER THERMAL POWER
- b. Shutdown D.d0% o,0 ti%
' < t-798% of RATED < 0-89 A o'f RATED THERMAL POWER THERMAL POWER C. Core Protection Calculator System
- 1. CEA Calculators Not Applicable Not Applicable
- 2. Core Protection Calculators Not Applicable Not Applicable D. Supplementary Protection System Pressurizer Pressure - High 5 2409 psia 5 2414 psia II. RPS LOGIC A. Matrix Logic Not Applicable Not Applicable B. InifiationLogic Not Applicable Not Applicable III. RPS ACTUATION DEVICES s
A. Reactor Trip Breakers Not Applicable Not Applicable B. Manual Trip Not Applicable Not Applicable
TABLE 2.2-1 (Continued)
.: -REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMI TABLE NOTATIONS (1)
Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypas shall be automatically removed when THERMAL POWER is less than or equal to 10 4% of RATED' THERMAL POWER.
-::-r-(2)
In MODES 3-4, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided*the margin between the pres-surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be ' manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia. (3) In MODES 3-4, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached. y-(4) % of the distance between steam generator upper and lower level wide range instrument nozzles. b k (5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncer-tainties, and dynamic allowances. Trip may be manually bypassed below.tLIO + of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 'N.of RATED THERMAL POWER. tu't% The approved compensation. DNBR limit is 1.231 which includes a partial rod bow penalty If,the fuel burnup exceeds triat for which an increased rod bow penalty is required, the DNBR limit shall be adjusted. In this case a DNBR trip setpoint of 1.231 is allowed provided that the difference is com-pensated by an increase in the CPC addressable constant BERR1 as follows: RB - RB ' o d (% POL) BERR1 = 8 RR1 +
- new old 100 d (% DNBR) 3 where BERRi old is the uncompensated value of BERR1; R8 is the fuel rod bow penalty in % DNBR; RB, is the fuel rod bow penalty in % DNBR already accounted for in the DNBR limit; POL is the power operating limit; and d (% POL)/d (% DNBR) is the absolute value of the'e most adverse derivative of POL with respect to DNBR.
PALO VERDE - UNIT fl 2-5
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4 REACTIVITY CONTROL SYSTEMS 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL 46 G4s R.ur wsc~.erdD SHUTOOWN MARGIN - T., OREATER-THAN-210*F-LIMITING CONDITION FOR'0PERATION 3.1.1.1 .
/.
delta k/k.The SHUTDOWN MARGIN shall be greater than or equal to g.0% APPLICABILITY: MODES (( 3? end 474vo 5 ard ,4a #wd-de~ucr# ACTION: #5 V #MM'O
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WiththeSHUTDOWNMARGINlessthan,6(0%deltak/k,immediatelyinitiateand
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continue boration at greater than or equal to 26 gpm to reactor coolant system of a solution containing greater than or equal to 4000 ppm baron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REOUIREMENTS F l L 4.1.1.1.1 to 6'.0% delta Tiiek/k: SHUTDOWN MARGIN shall be determined to be greater than or eq
/ .' .'~5/ithin 1 hour after detection of an inoperable and atCEA(s) least 3
on6ger 12 hours thereafter while the CEA(s) is inoperable. ,,,Jf the inoperab'l'e-CEA i,s immovable as a result of excessive,fr.icti~en or mechanical interfirence_or known to be untrippable,'the above re-
; quired SHUTDOWN MARGIN sh'all-be,yerified. acceptable with an increased i allowance for the withdrawn worth'of the immovable or untrippable CEA(s). ./~ .,
lb. When in MODE-I'oIMODE 2 with K ~ greater than or equal to 1'.0,,at
- iL-j east'6i1ce per 12 hours by verifyfng that CEA group withdrawal ~is 's -
within the Transient Insertion L.imits_,of_ Specification 3.1.3.6.
]
/ 00E-2 w hK ~
achieving reacto (r cri dbaless than 1.0, with' or to critical CEA ' ng that the predicted thin the Tiiitts of-Speqification 3.1.3.6.
" " See Special Test Exception 3.10.J.';*
e-
.PALO VERDE - UNIT l 3/4 I-1 l , ,
,g,, o m-. - -
SURVEILLANCE REQUIREMENTS (Continued) id. Prior to inii,iel vperation_above 5%. RATED-THERMAL 70WER a'fter each fuelJadingdy consitfe'Fhti6'nifWrattors-of-e below,.with the
..-CEA groups at the Transient Insertion Limits of Specification'rl:3.6s n,
ce s-
& "f.
When in MODE 3.e6 4,^at least once per 24 hours by consideration of at least the.following factors:
- 1. React 6r' Coolant System boron concentration,
- 2. CEA position, 3.
Reactor Coolant System average tempdrature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
, 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted
{ values to demonstrate agreement within + 1.0% delta k/k at least once per
\ 31 Effective Full Power Days (EFPD). This compar.ison shall consider at least g those factors stated in Specification 4.1.1.1.17 above. The predicted *\
reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. [h k LLu.h LD CLC s' O,
,PALOVERDE-UNITfj , 3/' -i t .i . ~ .
. l. .. . $
. . . .. b?.
- .. ._ . . _ _ . _ - _ _ - - _ _ . - _ . ' - [ Y & $ U .' x .-
REACTIVITY CONTROL SYSTEMS C _SHUTOOWN MARGIN gy.,-gy f CdM M YN.O? M '
-T g L4SS-THAN-OR-EOUAL-TO-210*F-
_ LIMITING CONDITION FOR OPERATION P ,- s3.1.1.2 delta k/k.The SHUTOOWN MARGIN shall be greater than or equal to _-4.0% '
\ - _ . ,,
APPLICABILITY: MODE 5. 5(c' M# #' ACTION: With the SHUTDOWN MARGIN less than 4.0% delta k/k, immediately initi continue boration.at o greater than or equal to 26 gpm to the reactor coolant system of a soluti' n containing greater than or equal to 4000 ppm baron or equivalent until the' required SHUTDOWN MARGIN is restored. j
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' l i SURVEILLANCE RE0UIREMENTS N -' I
( ,
;4.1.1.2 to 4.0% delta The
/
( k/k: SHUTDOWN MARGIN shall,be determi N
; a. i Within 1 hour after d'etection of an inoperable CEA(s) and at I least once per 12 hours thereafter while the CEA(s) is inoperable.
If the inoperable CEA is immovable as a result of excessive friction
, or mechanical interference or known to be untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the / withdrawn worth of the immovable or untrippable CEA(s).
l b. At least once per 24 hours by consideration of he,following factors,:
- 1. R
/ eactor Coolant System boron concentration,
- 2. ~ CEA position, 3i Reactor Coolant System average temperature,
. 4. '
t
/ 5. Fuel Xenonburnup based onand concentration, gross thermal energy generation, 'N, i
- 6. Samarium concentration. ',
. l
\
~ .
\
PALOVERDE-UNITf( 3/4191
.t - '
NE.Ad ( ( Me -- AEACTIV!?Y CONTROL SYSTEMS \ SHUTCCVN MARGIN - Kn.1 - ANY CEA WITHDRAWN LIMITING CONDITION FOR OPERATION z 3.1.1.2
- a.
The SHUTCOWN' Figure 3.1-1A. and MARGIN shall be greater than or equal to that shown in b. For Tcold less than or equal to 500 0F, K N -1 shall be less than 0.99. APPLICABILITY: MODES 1, 2*, 3*, 4*, partially withdrawn. and 5* with any full-length CEA fully or ACTION: a. With the SHUTDOWN MARGIN less than that in Figure 3.1-1 A, immedi initiate and continue boration at greater than or equal to 26 gpm to the reactor coolant system of a solution containing greater than or equal to 4000 MARGIN ppm ofand is restored, boron or eouivalent until the required SHUTDOWN b. With Teo14 less than or equal to 500 0F and NK -1 greater than or equal to 0.99, immediately vary CEA positions and/or initiate and continue boration at greater than or ecual to 26 gpm to the reactor coolant baron or equivalent until the required Ksystem of a solution contai N-1 is restored. SURVEILLANCE RECUIREFENT3 4.1.1.2.1 With any full-length 'CEA fully or partially withdrawn, the SHUTDOWN M shall be determined to be greater than or equal to that in Figure 3.1.lA: a. Within 1 hour after detection of an incoeraola CEA(s) and at least once per 12 hours thereafter wnile tne CIA (s) is incoeracle. If the ine:eracle CIA is immovaele as a result of excessive friction or meenanical interference or known to me untrip;aele, the ateve required SHUTCOWN MARGIN snall be increased by an amount at least equal to the witnerawn worth of the immovaale or untrippa le CEA(s). See Special Test Exceptions 3.10.1 and 3.10.9. PALC VERDE - UNIT 1 3/41-2
- n. - '
. m t,Y 1 m \ n s V ,
o t G~ %
.s.
7 r-
......__._.i.__-,1~....____..I__...,_.....__!______.----
S.
. . . - : .......t _.__t..._....
. _. .. ... _. __ __ ____.__.._.f..___
. .. .. ,+
. _ ... . .._.___. _.____ _ . . . .. _ _ __ _ i 6 _
~
- . .;. ... . ._..; _. _.._- w___ e_
i t
= . _ _
........_.;... i 'sM s <'iu ',
.._.. .. _ ._._. p ... . ..._- ..;
._....__._._....._..j......._.
..J______.__....._4
..._.p. i. .. . .
_.._..._....s.__. t . - - .. . .. v__..._-._... 4._. 5 _ _ . . .
...._...._.a....__ _ . . _ . . . _ . . . . _ . . . . . _ . ._ .-_. . . .
s..__.... p_ . m w . . . . . _ .. w ._ __.._- EEG ION OF ------ -- . _ _ .-.-.-.- - __.. . ...;.._.___. ACCEPTABLE .. r g . _ . . .. _. ._._. _. _. . _
. OPERATION
~~
~~
r
. . .. . _ . . ' ~ .___ ~
. . ~ ' .-
t-. 4 ' a . . . W .. A .. ., , . n .._.._..._-..;...__..... _ . . . _ _ _ . . . . m ..
== . . .
. ._1.
M .. e 3 s,.
. [EGIOM~0r...
- n .
. - UNACCEPTA.BL. E..- -- 1 C . . . .
. _ . .OP.ERATION .
B _.
.!.2 -
M 2 _ . _ _ . ... 1
-- '(-21071-) . . . . _ . . . . . . . _ ,
. .. . _ . . ,i O 100 200 300 400 500 600 COLD LEG TEMPERATURE (*F)
FIGURE 3.1 - 1A SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE PALO VERDE - UNIT 1 3/4 1-2a I .
hed g (, -] REAC- */! Y CON 7tCL SYSTius . SURVi!CLNCE RECU!XEMENTI /C:ntinuec) .
' ' ' . . . .h . When
leas in MCCE 1 cc MCCE 2 with K ,, greatar ::an or ecual ::
1.0, at enca per 12 neurs by verifying sna: CIA gr:u: wita:rawal is witnin :ne- Transient Inserti:n Limits of Specificatica 3.1.3 ..5
. . c.
When in MCCE 2 with K,,, less snan 1.0,, witnin a hcurs Orier t= acnieving rete :r critidality by verifying na :ne =recic ac critt:al CIA positi:n is witnin :na limits of Specificati n 3.1.3.5.
- d. , Price ::
d initial c:gratien a:cve 5% RATED THERMAL PCWER aftar eac CEAuel greu loading, by c:nsideration s at tne Transian: of :ne fact:rs of e. telcw, wita :ne e. Inserti:n Limits of 5:ecifi:ati:n 2.1.3.5. When at leas:insta McCE 3, 4, fact:rs: f:llcwing or 5,at leas ence oer 24 hours by censiderati 1.
- 2. React:r CIA C: clan: Systac :cr:n ::ncan:ra-f n,
;csitien, 3.
4 React:r C: clan: Systam average tam:erature Fuel turnu: =as
- 5. Xenen ::ncan:ra:ac f:n, nanc gr:ss narmal energy g,enerati:n, 6.
SE:arit: c:ncantraticn. 4.1.1.2.2 When in and withdrawn, 1100E Tcold 3, 4, or 5, with any full-length CEA fully or ocr less than or equal to 500 CF, KN -1 shall be determined to be less than 0.99 at least once per 24 hcurs by consideration of at least the follcwinn factors: 1.
- 2. React:r CEA C: elan Systam :ren ::ne'entratten,
- siti:n,
- 3. React:r C::lan: Systa 4 Fuel :urnus :asec :n :gr:averags tas:erature, ,
- 5. Xencn c:nenntra:f=n, andss ::ar=al energy generati:n, -
6. Samarium ::nesn:ra:i:n. 4.1.1. 2. 3 The overall c:ra reactivity balanca snall be c:::arec :: values 31 Ef" active : cemenstrata Full Fewer Daysagreenent(EFFO). witnin - 1.C% celta k/k at leas; enc =r a :er reac-ivity values snall te adfustac (nerealt:ac) t:the c:re c:nditiens prier :: e leacing. ex aecing a fuel turnup of EO EFF0 aftar ea:n fuelc:r PAL 3 VERCE - UNIT 1 3/4 1-3' i . l . .
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 be OPERABLE: At least two of the following three boron injection flow paths shall a. A gravity feed flow path from either the refueling water tank or the spent fuel pool through CH-536 (RWT Gravity Feed Isolation Valve) and a charging pump to the Reactor Coolant System, b. A gravity feed flow path from the refueling water tank through CH-327 (RWT Gravity Feed / Safety Injection System Isolation Valve) and a charging pump to the Reactor Coolant System, c. A flow path from either the refueling water tank or the spent fuel pool through CH-164 (Boric Acid Filter Bypass Valve), utilizing gravity feed and a charging pump to the Reactor Coolant System. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT( STANDBY and-borated-to-a-SHUTDOWN-MARGIN-equivalent-to-at-4 east-6%-delta-k at-210 F-within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REOUIREMENTS 4.1.2.2.1 OPERABLE: At least two of the above required flow paths shall t,e demonstrated
- a. .
At least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path that b. At least once per 18 months when the Reactor Coolant System is at normal operating pressure by verifying that the flow path required by Specification 3.1.2.2 delivers at least 26 gpm for 1 charging pump and 68 gpm for two charging pumps to the Reactor Coolant System. 4.1.2.2.2 ' Mode 3 or Mode 4 to perform the surveillance testing of provided the testing is performed within 24 hours af ter achieving normal operating pressure in the reactar coolant system. 3/4 148 PALOVERDE-UNIT [{
.t - -
_ CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. _ APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: _g. With only one charging pump OPERABLE, restore at least two charg OPERABLE SHUT 00WN-MARGIN status equivalent-to-at-least within 72 hours or6%_ be delta in at.k/k least
-a-HOT S at-210?F-w 6 hours; restore at least two charging pumps to OPERABLE status wi 7 days or be in COLD SHUTDOWN within the next 30 hours.
SURVEILLANCE REQUIREMENTS ( 4.1.2.4 by Specification 4.0.5.No additional Surveillance Requirements other than those e. PALOVERDE-UNITf( . 3/4 1-10
.'\. -
BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 Each of the following borated water sources shall be OPERABLE:
- a. The spent fuel pool with:
1. A minimum borated water volume as specified in Figure 3.1-2, and 2. A boron concentration of between 4000 ppm and 4400 ppm boron, and
- 3. A solution temperature between 60 F and 180*F.
- b. The refueling water tank with:
- 1. A minimum contained borated water volume as specified in Figure 3.1-2, and 2.
A boron concentration of between 4000 and 4400 ppm of boron, and
- 3. A solution temperature between 60 F and 120 F.
APPLICABILITY: MODES 1, 2,* 3,* and 4*. ACTION: a. With the above required spent fuel pool inoperable, restore the pool to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours er.dhrated te a SHUT 00',JN MARGIN qsholeni.-
)rk Ln Ica n-6L ocito L'ket210-Fdestoretheaboverequiredspentfuelpoo to OPERABLE status within the next 7 days or be in COLD SHUT 00WN within the next 30 hours.
b. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.6 OPERABLE: Each of the above required borated water sources shall be demonstrated
- a. At least once per 7 days by: '
- 1. Verifying the baron concentration in the water, and 2.
Verifying the contained borated water volume of the water source.
- b. At least once per 24 hours by verifying the refueling water tank temperature when the outside air temperature is outside the 60*F to 120"F range, c.
At least once per 24 hours by verifying the spent fuel pool temperature when irradiated fuel is present in the pool. See Special Test Exception 3.10.7. l
- 3/4 1-13 PALOVERDE-UNIT ;
{(
TABLE 3.3-1 (Continued) , REACTOR PROTECTIVE INSTRUMENTATION TABLE NOTATIONS
*With the protective system trip breakers in the closed position, the C drive system capable of CEA withdrawal, and fuel in the reactor vessel.
#The provisions of Specification 3.0.4 are not applicable.
- :=--
(a) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER bypass shall be automatically removed when THERMAL POWER is less; equal to 10 4% of RATED THERMAL POWER. (b) Trip may be manually by)assed below 400 psia; bypass shall be automatically equal removed wienever pressurizer pressure is greater than or to 500 psia. tod% (c) Trip may be manually bypassed below 1% of RATED THERMAL POWER bypass shall be automatically removed when THERMAL POWER is gr; ea or equal to 2% of RATED THERMAL POWER. Im% (d) Trip ma 3.10.3.y be bypassed during testing pursuant to Special Test Exception See Special Test Exception 3.10.2. I C ((e) f) There are four channels, each of which is comprised of one of the four reactor trip breakers arranged in a selective two-aut-of-four configuration (i.e. , o,ne-out-of-two taken twice). ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than require the Minimum Channels OPERABLE requirement channel to OPERABLE status within 48 hours, or be in at leastre HOT system STANDBY within the next 6 hours and/or open the protective trip breakers. ACTION 2 - . With the Number number of Channels of channels OPERABLE one less than provided the inoper,able channel is placed in the byp tripped condition within 1 hour. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with Saecification 6.5.1.6.g. The channel shall be returned to 0)ERABLE status no later than during the.,next COLD SHUT 00W PALO VERDE - UNIT / l.. 3/4 3-5
- 3. .. - '
3/4.10 SPECIAL TEST EXCEPTIONS N 3/4.10.1 SHUTDOWN MARGIN App 4., - dC4 A&R,W ec.i. 73 LIMITING CONDITION FOR OPERATION WKm -r 3.10.1 The SHUTDOWN MARGINarequirementofSpecification3.1.1.Jmaybe suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLEnCEA(s), or the reactor is subcritical by at least the reactivity equivalent of the highest CEA worth. APPLICABILITY: MODES 2, 3* and 4*#. ACTION: a. With any full-length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immedi-ately initiate and continue boration at greater than or equal to 26 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN'/ required by Specification 3.1.1./grestored. MO Kv-i b. With all full-length CEAs fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 26 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. C SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length and part-length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTOOWN MARGIN to less than the limits of Specification 3.1.1.1. 4.10.1.3 When in MODE 3 or i!0DE 4, the reactor shall be determined to be subtritical by at least the reactivity equivalent of the highest estimated CEA ' worth or the reactivity equivalent of the highest estimated CEA worth is avail-able for trip insertion from OPERABLE CEAs at least once per 2 hours by con-sideration of at least the following factors:
.a. Reactor Coolant System baron concentration,.
- b. CEA position,
- c. Reactor Coolant System average temperature, 5
- d. Fuel burnup based on gross thermal energy generation,
- e. Xenon concentration, and
- f. Samarium concentration.
Operation in MODE 3 and MODE 4 shall be limited to 6 consecutive hours. Limited to low power PHYSICS TESTING at the 320*F plateau. PALOVERDE-UNIT /{ 3/410'-{0
, f, uVf . -
3/a.10 SPECIAL TEST EXCEPTIONS 3/4.10.9 . SHUTUC'4N MARGIN AND Xg.]
. CEDMS TESTING LIMITING CONDITTON FOR OPERATION 3.10.9 SHUTDOWN MARGIN and K -1 requirements N
for pre-startup drive tests mechanism system to demonstrate provided: the OPERABILITY of the con suspended element
- a. No more than one CEA is withdrawn at any time,
- b. No CEA is, withdrawn more than 7 inches,
- c. the Thestart KN.1 requirement of Specification 3.1.1.2 is met prior of testing,
- d. suspended All otherduring operations involving positive reactivity changes the testing.
APPLICABILITY: MODES 4 and 5. ACTION: With any of the above requirements not met, suspend testing a comply with the requirements of Specification 3.1.1.1 or 3.1.1.2, . SURVEILLANCE RECU!?!MENTS 4.10.9 within one hour prior to the start of testing, and a during testing. s PALO VERCE - UNIT 1 . 3/4 10-9
\- *
" " s s.
,. .. CONTROLLER BY USER ( SPECIAL TEST EXCEPTIONS . c 3/4.10./ NATURAL CIRCULATION TESTING PROGRAM
- LIMITINGCatIDITIONFOROPERATIOi IO 3.10.5l The limitations of Specifications 3.4.1.2, 3.4 1 3
. . , and 3.7.1.6 may be suspended Program during
- provided: thg_ performance of the Startup Natural Circ a.
Operations Coolant System areinvolving a suspended. reduction in boron concentration of b. Core outlet temperature. temperature is maintained at least 10*F below Sa c. A reactor coolant pumo shall not be started 255 F during cooldown, or 295 F during heatuo, unless the water temperature (saturation temperature corresponding to steam each of the Reactor Coolant System cold leg tempe APPLICABILITY: MODES 3 and 4 during Natural Circulation Testing. ACTION: With the Reactor Coolant System saturation ma pump. SURVEILLANCE REOUIREMENTS IO 4.10.p.1 The saturation margin shall be determined to be within the above by Table 3.3-10 or, by calculating the saturation 30 minutes.
*Startup Natural Circulation Testing Program:
Natural Circulation Cooldown Test at 80% power. PALO VERDE - UNIT 1 3/410-fl0 Amendment No. 2 e i mNTmii m nv' ucro -
3/4.1 REACTIVITY CONTROL SYSTEMS _ BASES. - ) 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN /tMD K ., _. _- g i subcritical from all 'opeFating conditions, (2) the reac j able limits assuming the insertion of the regulatin limits of Specification 3.1.3.6, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. , x. Mr REMSEP WRWEMG , i N. neg.f- ps
- fuel depletion, RCS boron concentration, and RCS TSHUTOO
{conditionoccursatEOL,withT eof [3 The most restrictive at n 1 ad.op'erating temperature, and is
! cold ) associated trolled RCS cooldown.with a postulated steam line breik accident and resulting uncon- '
- MARGIN of 6.0% delta k/k is required'to control the reactivity t -
j r
; Accordingly, the SHUTDOWN MARGINequirement is '.
dependent Safety Analysis.CEA insertion. limits and with the assumptions used in the FSAR t N., , { With T \ i I from uncont , cold ess than or equal to 210 F, the reactivity transients' resulting j requirement rolled RCS is set cooldown to ensure thatare minimal reactivity and a 4% transients Ak/k SHUTDOWN resulting from an
@nadvertentsingleCEAwithdrawaleventareminimal.
{ 3/4.1.1.3 i MODERATOR TEMPERATURE COEFFICIENT (MTC) ' i The limitations on moderator temperature coefficient (MTC) are provided remain valid through each fuel cycle.to ensure that the assumptions use ! The surveillance requirements for
- i measurement of the MTC during each fuel cycle are adequate to confirm the MTC
! value since this coefficient changes slowly due principally to the reduction ! ( in RCS baron concentration associated with fuel burnup. , ! the measured MTC value is within its limit provides assurances that the coef-i ficient will be maintained within acceptable values throughout each fuel cycle. he (e.{-e (AcGe) & Dh 4 , PALOVERDE-UNITf1 . B 3/4 1.-l l 4 , . ,
hN kk k ( 3/A.1 4EAC7I'/I'? C'N770L 5'/57"5 ' BASES
~ _ . . _ . -
,l/4.1. I !CRA7?CN C*NTICL 3/1.1.1.I and 3/t.1.1.2 "!FUT C'4N m ; GIN - '
AND XN ,) --.-..- . .~. - . - suberitical following a design basis accident or anticipa occurrence. The functionNof K -1 is to maintain sufficient subcriticality to preclude inadvertent criticality following ejection of a single control eleme assembly (CEA). During operation in MODES I and 2, with keff greater than equal to 1.0, the transient insertion limits of Specification 3.1.3.6 ensure that sufficient SHUTDOWN PARGIN is available. SHUTCOWN PARGIN is the amcunt by which the core is subcritical, or would be suberitical immediately following a reactor trip, considering a malfunction resulting in the highest worth CEA failing to insert. is a Kg.1 measure of the core's reactivity, considering in the highest worth inserted CEA being ejected. a single malfunction resulting SHUTCCWN PARGIN requirements vary throughout the core life as a function (T of fuel depletion and reactor ccolant system (RCS) cold leg tem cold). The most restrictive condition occurs at EOL, with T,,,,ld at no-load operating temperature,and is associated with a postulated ste M sine break accident and the resulting uncontrolled RCS ccoldewn. In the analysis of t accident, the specified SHUTCCWN PARGIN is required to control th transient and ensure that the fuel performance and offsite dose criteria a satisfied. As (initial) Tcold resulting reactivity transient are less severe and, therefore, th SHUTDOWN MARGIN also decreases. Eelow T deboration event beccmes limiting wi_th respect to the SHUTCOWN MA requirements. Below 210 oF, the specified SHUTCOWN MARGIN ensures that sufficient time for operator actions exists between the initial indication o the deboration and the total loss of shutdcwn margin. Accordingly , with at least ' based upon these limiting conditions.one CEA partially or fully withdraw - PARGIN that are not limiting with respect to the Specif single CEA withdrawal and startup of an inactive reactor coolant pump. Kg.1 requirements vary with the amount of positive reactivity that would be introduced assuming the CEA with the highest inserted worth e the core. In the analysis of the CEA ejection event, the KN.] requirement ensures that the radially averaged enthalpy acceptance criterion is satisfied, considering power redistribution effects. Above Tcold of 500 0F Dopoler reactivity feedback is sufficient to preclude the need for a spe,cific X N-1 are equivalent in terms of minimum acceptable core bor PAL 3VE.7,0!.UNITi , 8 3/4 1 1,
- i. -
2/4.1 1E.sCTIV!T? C N770L SYSTim '
._.1 -
EASES 3/4.1.1 20 RATION C0lTROL (cont.) . 3/4.1.1.1 and 3/4.1.1.t'52UTCCW .uARGI.'l AND Xn.1 (ccat l.___. "
- Other technical specifications that refere[1ce the Specification SHUT 00MN MARGIN or N-1 K are:
CONTROL ASSEMBLIES, 3/4.9.1, REFUELING OPERATIO 3/4.10.1. SHUTCOWN MARGIN AND N X CECMS TESTING. MARGIN N AND X CEA WORTH .
. , SHUTCOWN TESTS, an 3/ , . ; .1. 3 ,y . .
. . a . . .z _ .s. . t.,,u. .--
. A ., .x-
- e. . ..-. . e....
. . . . ,1., . , y'.. 1
- The ca:
ensure limits fcas en =ccarat:r as:aratura c:affi:ian: ' remain valid thr:ugn eaca fuel cycle.ne asst ==:icns ansient usac in :na ac:i:an; a ar.alysis . The surveillanca escuiremen: 3 for value sinca this c:afficient cnangas s'culy cuene MTC
- na escuc ::n rinci: ally in RC5 the =easured ter:n MTCc:ncan:.atica value is wi:Min associa:ac wiO fuel . The turnu:
c ficien will ta =aintainac wit?.in 1::a:n:la valuas cycla. ca :ne ea:n fuel inr:ugn
- af- u::s lim 3/A.1.1.4 -
MINI.'"JM TEMPERA'"URE OR CRITICALIT? . . _ _ . . . . . .. _ .... This specificati n ensures sna: the reacter will ne the React:r C: clan Systa: c:1c leg taccarature less than 552*Fte ace critical with is required analy:sd temperature to ensure ran (1) tha =ccarat:r tam:erature c:afficient This limits:icn is witnin
;s ner:al c:erating range, ge, (2) the protactive instrumentatien is within its analysis. and (3) to ensure c:nsis:ancy wita :?.e FIAR safa:y .
4 1 PALO V!F.:E - UNIT 1 , a 3/4 1-la
. l -
_ . - - - - -- ~ . - - - -. - --- -. - - - -- d 1 REACTIVITY CONTROL SYSTEMS BASES i I 2 3/4.1.1 4 MINIMUM TEMPERATURE FOR CRI ITY the Reactor Coolant System cold leg temperature less tha This ifmitation analyzed temperature ~ range, (2) the protective ins normal operating range, Sand (3) to ensure consistency with the FSAR safe analysis. W ). & ' { j 3/4.1.2 BORATION SYSTEMS f pggg g g[ {
- available during each mode of facility operation.The boron injectio The components required to perform this function include (1) borated water sources, (2) charging pumps, 1 (3) separate flow paths, (4) an emergency power supply from OPERABLE diese i
capable of isolating the VCT from the charging pump suction l The nominal i capacity of each charging pump is 44 gpm at its discharge. Up to 16 gpm of this may be diverted to the volume control tank via the RCP control bleedoff.
- yielding the 26 gpm value. Instrument inaccuracies and pump performance 1
With the RCS temperature above 210*F a minimum of two separate and redundant boron injection systems are prov,ided to ensure single functional g
- capability in the event an assumed failure renders one of the systems inoper-able.
\
Allowable out of service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility } safety from injection system, failures during the repair period. i puc:a Srs;2M is G/N8ci- oi~ MtVoM4 Bd4;Ocs)dCVM~'W/7' 70 A
.The be-e+ica SHUTOOWN MARGIN ktmr-expectcapahi.lity oLaither-system..is_ sufficient to provid i
decay and cooldown to 210*F. operating conditians-of 4% delta k/k after xenon ! The maximum expected boration capability require-i ment occurs at EOL from full power equilibrium xenon conditions and requires 23,800 or gallons the spent fuelofpool. 4000 ppm borated water from either the refueling water tank
. Q.st (se au-@ .
With the RCS temperature below 210*F one injection system is acceptable } without single failure consideration on the basis of the stable reactivity l i condition of the reactor and the additional restrictior.s prohibiting CORE l i ALTERATIONS system becomes inoperable.and positive reactivity changes in the event the single injection ! The restrictions of one and only one operable charging pump whenever reactor coolant level is below the bottom of the pressur-i izer is based on the assumptions used in the analysis of the boron dilution event, j I '. Y y a h a na d i $ + j ! - A--TAe-boron-capabiUty-required below.210 F-is. based upon providing a 4% ' l ' delta-k/k-SHUTDOWN. MARGIN.af ter-xenon-decay.and cooldown.from 210"F to 120 This condition requires 9,700 gallons of 4000 ppm borated water from either the refueling water tank or the spent fuel pool. PALOVERDE-UNITf( . B 3/4 1-2 j . 6 .
irtserY sg 7xaNnwr, rwamr/en cssoci7Y ofene>9 srsren is xsnr 7;wn sa:p=7 car 70 s4;2s#v Tx . SWuTDxdAJ M4At/AJ 4%D/ce Ko , iQnnwnbw& VIE Of" Xt2 ,SProA~cA770xAS. . hI&Cl' Y2 fACH s737&"Af /3 CdB4Bdc' OkPAbMD/LiG 88472s4.) rawarieur 70s surce.uu meew cs 4% Da27A k/c. T//ER& enc: 7W 894.wD cxf>rc/7 v 0/:~ 7//f SYS7cw RfGM?n4D Ed20<d E'sC '/' ~ /3 mal"
~
7/&fL' Sc.>%C2~&7~ 7C~ S471SJ=~Y Tha~ SMO)Zbo.& P.4AWMJ A&D,kR K,y-, Attcw/MA2se.ns Op ;wer i seec/acs r w g. e e ee e-2 o e . . .. . \ . z. . . : . - -
3/4.10 SPECIAL TEST EXCEPTIONS BASES i 3/4.10.6 SAFETY INJECTION TANKS injection system check valves.This special test exception permits testing the The pressure in the injection header must be through the check valves; reduced below the head ,of the low pressure injection must be closed in order to accomplish this.The safety injection tank (SIT) isolation The SII isolation valve is still capable capabilityofshould automatic not operation in the event of an SIAS; therefore, system be affected. 3/4.10.7 SPENT FUEL POOL LEVEL spentThis fuel special test exception permits loading of the initial core with the pool dry. 3/4.10.8 SAFETY INJECTION TANK PRESSURE This special test exception allows the performance of PHYSICS TESTS at low pressure / low temperature (600 psig, 320 F) conditions which are required - to verify the low temperature physics predictions and to ensure the adequacy of design codes for reduced temperature conditions. 3/4 /O. ? SkurCou/A/ ManGir/ .dA/D Mt '-/ - CELW5 ?237/ 1'v!S iM.wt r.37~ E,VGPrior/ A;dcwS &&~ FeRADFM4A./Cf of CCA/TROc fcfyrs)7~ p/?/g,'g* Afgag%'Sp? s 725TS n?/OR 7& 37~4?rc/P(a,yrsvotif* Tiff 09fB'70R /H&%G 7D Bf CC^t*4;274fD 45 ro g,urmE9 SPECRo?7)od 3.//./ cR 3.J.i. 2 iS APPdKWE&c"
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ce/pcastrr ws/MG Tws t=5T*. PALOVERDE-UNITpl 8 3/4 10-2
. a .
INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 1.2 ACTI0N....................................................... 1-1 AX I A L SHAP E IND EX. v. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-1 ......... 1.3 AZIMUTHAL POWERq TILT - T ................................... 1-1 1.4 CHANNEL CALIBRATION......................................... 1-1 1.5 CHANNEL CHECK............................................... 1-1
- 1. 6 CHANNEL FUNCTIONAL TEST..................................... 1-2
- 1. 7 CONTAINMENT INTEGRITY....................................... 1-2 1.8 CONTROLLED LEAKAGE............. .................. ......... 1-2
- 1. 9 CO R E A LTE R AT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-2 .........
1.10 DOSE EQUIVALENT I-131....................................... 1-3 1.11 E - AVERAGE DISINTEGRATION ENERGY........................... 1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME.................... 1-3 I 1.13 FREQUENCY N0TATION.......................................... 1-3 1.14 GASE0US RADWASTE SYSTEM..................................... 1-3 1 15 IDENTIFIED /% Xy i 4
' - LEAKAGE.......................................... 1-3
> /-4
- 1. lfr7 MEMBER (S ) 0F THE PUB LIC. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 .....
1.1/00FFSITEDOSECALCULATIONMANUAL(0CDM)...................... 1-4 1.1#? OPERABLE - OPERABILITY...................................... 1-4
- 1. E.'.00 P ERATI ON A L MOD E - M0 D E . . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 ..........
- 1. go / PHYS I C S T E S TS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.2f2 PLANAR RADIAL PEAKING FACTOR - F 1-4
~
xy..........................
- 1. 2,23 PRESSURE BOUNDARY LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1,4,5' 1.274 PROCESS CONTROL PROGRAM (PCP) . . . . . . . . . . . . . . . . . . . . . .1-5 .........
- 1. 24'iP U R G E - P U R G I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 ..............
- 1. 25'6 RATED TH E RMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 .......
1.26'7 REACTOR TRIP SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . 1-5 ..... 1.278 REPORTABLE EVENT............................................ 1-5 1.28? SHUTDOWN MARGIN............................................. 1-fd 1.29NSITEB00NDARY............................................... 1-6 1.30'/ 50FTWARE........................................... ...... 1-6 PALO VERDE - UNIT 2 I 4 .
j INDEX DEFINITIONS SECTION PAGE 1.311 SOLIDIFICATION.............................................. 1-6
- 1. 373 SO URCE CHEC K. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6............
- 1. 3I4 STAGGERED TEST BASI S. . . . . . . . . . . . . . . . . . .1-6 ..........
1.3/fTHERMALP0WER............................................... 1-6 1.3J4 UNIDENTIFIED LEAKAGE........................................ 1.6' 7 l 1. 367 UNRESTRICTED AREA. . . . . . . . . . . . . . . . . . . . . . . . . . . .1-6 . . .7. . . . . . . . . 1.3.78VENTILATIONEXHAUSTTREATMENTSYSTEM........................ 1-7 1.3,8d i VENTING..................................................... 1-7 i i i 6 i k-4 2 4 i i i m i PALO VERDE - UNIT 2 II . L. -
4 INDEX f LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS ( SECTION PAGE 3/4.0 APPLICABILITY.............................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL 4 . g,g gg7gp SHUTDOWN MARGIN T 210 Fr 3/4 1-1 gy . m c. . . . . .a. . .umam .............. SHUTDOWN MARGIN - T g 210'F ....................... 3/4 1.3' Z. MODERATOR TEMPERATURE COEFFICIENT. . . . . . . . . . . . . . . . . 3/4 . . . .1-4 MINIMUM TEMPERATURE FOR CRITICALITY. . . . . . . . . . . . . . . . . .3/4 . 1-6 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTD0WN................................. 3/4 1-7 F LOW P ATH S - 0 P E RAT I NG. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4
.. 1-8 CHARGING PUMPS - SHUTD0WN............................. 3/4 1-9 CHARGING PUMPS - 0PERATING............................
3/4 1-10 BORATED WATER SOURCES - SHUTD0WN...................... 3/4 1-11 f' BORATED WATER SOURCES - 0PERATING..................... 3/4 1-13 I q, BORON DI LUTION A LARMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA P0SITION.......................................... 3/4 1-21 POSITION INDICATOR CHANNELS - OPERATING............... 3/4 1-25 POSITION INDICATOR CHANNELS - SHUTDOWN........... 3/4 1-26 CEA DROP TIME......................................... ..... 3/4 1-27 SHUT 00WN CEA INSERTION LIMIT.......................... 3/4 1-28 REGULATING CEA INSERTION LIMITS....................... 3/4 1-29 9 PALO VERDE - UNIT 2 IV.
- l.
- INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRE
_SECTION PAGE ELECTRICAL POWER SYSTEMS (Continued) 3/4.8.2 D.C. SOURCES OPERATING............................................ 3/4 8-9 3/4.8.3 SHUTD0WN............................................. 3/4 8-13 ONSITE POWER DISTRIBUTION SYSTEMS . OPERATING............................................ 3/4 8-14 3/4.8.4 SHUTD0WN............................................. 3/4 8-16 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES................................. 3/4 8-17 MOTOR-0PERATED AND BYPASS VALVES THERMAL OVERLOAD PROTECTION DEVICES................................. 3/4 8-40 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION..................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION......................................... 3/4 9-2 C 3/4.9.3 DECAY TIME.............................................. 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.....................r. ~3/4 9-4 3/4.9.5 COMMUNICATIONS.......................................... 3/4 9-5 3/4.9.6 REFUELING MACHINE....................................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P0OL BUILDING. 3/4 . . . 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL..................................... 3/4 9-8 LOW WATER LEVEL...................................... 3/4 9-9 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM................ 3/4 9-10 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES.............. ....................... 3/4 9-11 CEAs......................... ....................... 3/4 9-12 3/4.9.11 WATER LEVEL - STORAGE P00L.............................. 3/4 9-13 3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM.............. 3/4*9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MARGIN. 4.o. .....n. A:a. ?. . .Cc:M ivo?rd r< st.5
........................... 3/4 10-1 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS............. 3/4 10-2 3/4.10.3 REACTOR COOLANT L00PS................................... 3/4 10-3 %
- PALO VERDE - UNIT 2 IX .
1 . .
i _INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE................ 3/4 10-4 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY........ 3/4 10-5 3/4.10.6 SAFETY INJECTION TANKS.................................. 3/4 10-6 3/4.10.7 SPENT FUEL POOL 3 LEVEL.....................'.............. 3/4 10-7
.5p/4.10.j 10., SAFETY INJECTION TANK PRESSURE........
sontWN M.<kGW. %2? km , - ccan. . . . .s. re . . . s. .nuc ....... 3/4 10-8 3/4.11 RADIOACTIVE EFFLUENTS 9y 9 3/4'11.1 SECONDARY SYSTEM LIQUID WASTE DISCHARGES TO ONSITE EVAPORATION PONDS CONCENTRATION........................................... 3/4 11-1 00SE.................................................... 3/4 11-5 LIQUID HOLDUP TANKS..................................... 3/4 11-6 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE............................................... 3/4 11-7 00SE - NOBLE GASES...................................... 3/4 11-11 DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE F0RM..................... 3/4 11-12 GASEOUS RADWASTE TREATMENT.............................. 3/4 11-13 EXPLOSIVE GAS MIXTURE................................... 3/4 11-14 GAS STORAGE TANKS....................................... 3/4 11-15 3/4.11.3 SOLID RADI0 ACTIVE WASTE................................. 3/4 11-16 3/4.11.4 TOTAL 00SE.............................................. 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... 3/4 12-11 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... 3/4 12-12 PALO VERDE - UNIT 2 X .
\. - -
INDEX BASES SECTION PAGE 3/4.9.6 REFUELING MACHINE....................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P0OL BUILDING. B 3/4. . 9-2 3/4.9.3 SHUTDOWN COOLING AND COOLANT CIRCULATION. . .B. 3/4 . . . 9-2 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATIONSYSTEM................ B 3/4 9-3 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTO STORAGE POOL . . . . . . . . ........................
. . . . . . .. . . . R VESSEL B 3/4and 9-3 3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM. . . .B. .3/4 . . .9-3 3/4.10 SPECIAL TEST EXCEPTIONS
,w g , -ccw Mv?TW ~
3/4.10.1 SHUTD OWN MARG I N .4 . . . . . . +. . . ..............
. . . . . . . . . . .B. .3/4. . 10-1 72?S 75 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS. . . . . . . . . . . . . . . .
B 3/4 10-1 3/4.10.3 REACTOR COOLANT L00PS................................... B 3/4 10-1 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE................ B 3/4 10-1 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY........ B 3/4 10-1 3/4.10.6 SAFETY INJECTION TANKS.................................. B 3/4 10-2 3/4.10.7 SPENT FUEL POO L LEVEL. . . . . . . . . . . . . . . . . . . . . . .B. 3/4 . . . 10-2 3/4.10.8 SAFETY INJECTION TANK PRE 55URE..........................B 3/4 10-2
%.10.9 si.orm.uu MfCG.w 4w 4., - cc.cns -rssrwg 3/4.11 RADI0 ACTIVE EFFLUENTS gggg 3/4.11.1 SECONDARY SYSTEM LIQUID WASTE DISCHARGES TO ONSITE EVAPORATION P0NDS....................................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADI0 ACTIVE WASTE................... .............
B 3/4 11-5 3/4.11.4 TOTAL D0SE.............................................. B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-2 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 PALO VERDE - UNIT 2 XIV. I .
.. =. . _ - . . . . . _- _--
N-YlW@ _,.s__ ~ -_ ~_ %- - -x 1.16 K _ - inserteN" fully withdrawn. full-length control element assembly ofa hi OEFINITION seestt(M Y._ M / #Q MEMBER (ST)OF THE PUBLIC 1.1[7 EMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include I employees of the licensee, its contractors, or vendors. Also excluded from ! this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for plant.recreational, occupational, or other purposes not associated with the 0FFSITE DOSE CALCULATION MANUAL (ODCM) - l' 1.1/6The OFFSITE DOSE CALCULATION MANUAL shall contain the c and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program. s OPERABLE - OPERABILITY 1.18?A system, subsystem, train, component, or device shall be OPERABLE or 1
! have OPERABILITY when it is capable of performing its specified function (s),
- and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are i required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE - MODE (( 1.192 tan OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive co'mbination of core reactivity condition, power level, and cold leg reactor coolant temperature specified in Table 1.2. j PHYSICS TESTS 1.2 d PHYSICS TESTS shall be those tests performed to measure the fundament 1 nuclear characteristics of the reactor core and related instrumentation and ] (1) described in Chapter 14.0 of the FSAR, (2) authorized unifer the provisions , of 10 CFR 50.59, or (3) otherwise approved by the Commission. 1 PLANAR RADIAL PEAKING FACTOR - F g 1.21'OThe PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane i average power excluding densityofofazimuthal the effects the individual tilt. fuel rods in a given horizontal plane, PRESSURE BOUNDARY LEAKAGE 1.223 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall. PALO VERDE - UNIT 2 1-4 1 . .
DEFINITIONS _ PROCESS CONTROL PROGRAM (PCP) ' 1.23'-1 The PROCESS CONTROL PROGRAM shall contain the provisions to assure that the SOLIDIFICATION of wet radioactive wastes results in a waste form with properties that meet the requirements of 10 CFR Part 61 and of low level radioactive waste disposal sites. The PCP shall identify process parameters influencing SOLIDIFICATNN such as pH, oil content, H2 O content, solids content, ratio of solidificatipn. agent to waste and/or necessary additives for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each waste type, based on laboratory scale and full-scale testing or experience. The PCP shall also include an identification of conditions that must be satisfied, based on full-scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and af low level radioactive waste disposal sites. PURGE - PURGING 1.2,4gPURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentra-tion, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. j RATED THERMAL POWER L 1.2% RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3800 MWt. REACTOR TRIP SYSTEM RESPONSE TIME 1.2d7 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism. REPORTA8LE EVENT 1.2[6 A REPORTABLE EVENT shall be any of those conditions specified in Sections 50.72 and 50.73 to 10 CFR Part 50. SHUTDOWN MARGIN 1.2// SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. No change in part-length control element assembly position, and
- b. All full-length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.
PALO VERDE - UNIT 2 1-5 l .
DEFINITIONS SITE BOUNDARY 129NTheSITEBOUNDARYshallbethatlinebeyondwhichthelandisneithe owned, nor leased, nor otherwise controlled by the licensee. SOFTWARE 1.30'/ The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation, and procedures. SOLIDIFICATION 1.36SOLIDIFICATIONshallbetheconversionofradioactivewastesfromliquid systems to a homogeneous (uniformly distributed), monolithic,, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). SOURCE CHECK 1.3f3 A SOURCE CHECK shall be the qualitative assessment of channel respon when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.33k A STAGGERED TEST BASIS shall consist of: a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval. THERMAL POWER 1.3ffTHERMAL reactor coolant. POWER shall be the total reactor core heat transfer rat UNIDENTIFIED LEAKAGE 1.3I6 UNIDENTIFIED LEAKAGE shall be all leakage which does not constitute either IDENTIFIED LEAKAGE or reactor coolant pump controlled bleed-off flow, UNRESTRICTED AREA 1.36'?An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDA access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE B0UNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. PALO VERDE - UNIT 2 1-6 ,
DEFINITIONS C VENTILATION EXHAUST TREATMENT SYSTEM 1.3/6AVENTILATIONEXHAUSTTREATMENTSYSTEMshallbeanysystemdesigneda installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING 1.3dfVENTINGshallbethecontrolledprocessofdischargingairorgasfroma confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not
, imply a VENTING process.
C 9 9 PALO VERDE - UNIT 2 1 g , .
l l TABLE 2.2-1 (Continued) 5 l g REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS
% FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES (
- 2. Logarithmic Power Level - High (1)
- 0. c t0T, 0.011 q a. Startup and Operating < 9 998E of RATED < -0.-89 of RATED
" THERMAL POWER THERMAL POWER
- b. Shutdown O.00% o , 0 11 %
' < 4-798% of RATED < 0-895% of RATED THERMAL POWER THERMAL P0'WER C. Core Protection Calculator System
- 1. CEA Calculators Not Applicable Not Applicable
- 2. Core Protection Calculators Not Applicable Not Applicable D. Supplementary Protection System l Pressurizer Pressure - High < 2409 psia < 2414 psia II. RPS LOGIC
( A. Matrix logic Not Applicable Not Applicable i B. Initiation Logic Not Applicable Not Applicable l III. RPS ACTUATION DEVICES l A. Reactor Trip Breakers Not Applicable Not Applicable B. Manual Trip Not Applicable Not Applicable l l l l l
1 I i TABLE 2.2-1 (Continued)
- l. . . . - REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS ;
! TA8LE NOTATIONS 3 (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass l shall be automatically removed when THERMAL POWER is less than or equal to 10 4% of RATE 0' THERMAL POWER. 1 j j (2) In MODES 3-4, va'1 e may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pres-t surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer ! l pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed 1 whenever pressurizer pressure is greater than or equal to 500 psia. (3) In MODES 3-4, value may be decreased manually as steam generator pre.isure ' is reduced, provided the margin between the steam generator pressure and i . j this value is maintained at less than or equal to 200 psi; the setpoint ; shall be increased automatically as steam generator pressure is increased i l until the trip setpoint is reached. i 3 (4) % of the distance between steam generator upper and lower level wide i
- range instrument nozzles.
i
](k j (5) As stored within the Core Protection Calculator (CPC). Calculation of !
the trip setpoint includes measurement, calculational and processor uncer-t tainties, and dynamic allowances. Trip may be manually bypassed below 2%.109 - i of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL i POWER is greater than or equal to .I5'.of RATED THERMAL POWER. i id% l i The approved DNBR Ifmit is 1.231 which includes a partial rod bow penalty compensation. If the fuel burnup exceeds that for which an increased rod t' bow penalty is req,uired, the DN8R limit shall be adjusted. In this case a DNBR trip setpoint of 1.231 is allowed provided that the difference is com-j pensated by an increase in the CPC addressable constant BERR1 as follows: 1 i BERRi new
= BERRi old
+
100
- NB j where BERRi old is the uncompensated value of BERR1; R8 is the fuel rod
{ bow penalty in % DNBR; RB, is the fuel rod bow penalty in % DNBR already j accounted for in the ONBR limit; POL is the power operating limit; and d (% POL)/d (% DN8R) is the absolute value of the most adverse derivative { of POL with respect to DNBR.
- I 4
t 1 j ! i 1 .
,PALO VERDE - UNIT 2 2-5 ;
F
- l. .
REACTIVITY CONTROL SYSTEMS V4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.i BORATION CONTROL
- 4t.t, c4.9s R.a y' /45cW73*D SHUTOOWN MARGIN - T g CREATER-THAN-210*F-LIMITING CONDITION FOR~ OPERATION 3.1.1.1 '
/
delta k/k.The SHUTOOWN MARGIN shall be greator than or equal to /.0% APPLICABILITY: MODES (( 3? end 4fAvo S~ V ,4a' M7:# w 4-44 2 W ACTION: O ### With the SHUTDOWN MARGIN less than ,6'0% delta k/k, immediately initiate and continue boration at greater than or equal to 26 gpm to reactor coolant system of a solution containing greater than or equal to 4000 ppm baron or equivalent until the required SHUTOOWN MARGIN is restored. SURVEILLANCE REOUIREMENTS r I k . 4.1.1.1.1 to g.0% delta k/k:The SHUTOOWN MARGIN shall be determined to be greater t f Q...._ - -thin 1 hour after detection of an inoperable.-CEA
/.'a once-per 12 hours thereafter while the CEA(s) is inoperable.
1 inoperabl'e'CEAis immovable as a result of excessive,fr.iction.or If the mechanical inter'?hrence or known to be untrippable,'the above re-quired SHUTOOWN MARGIN sh'all-be yerified. acceptable with an increased allowance for the withdrawn worth'of"the immovable or untrippable CEA(s). ./' ., (b. When in MODE-Ior MODE 2 with X greater than or equal to T.'0,,at i east'o'nceper12hoursbyveri?NngthatCEAgroupwithdrawaiis ' L:___w3hin the Transient Inser,ti to3. L.imits.of_$_pecification 3.1.3.6. .,] 8
/. achievingODE reactor4cridbessK l t than 1.0, with_in_4 hours pMor to ertfy Ing that the predicted critical CEAW w thin thelTihits-of-Speg_ification i 3.1.3.6.
" " See Special Test Exception 3.10./.Y 3/4 I-1 -' #-
* .PALO VERDE - UNIT f l
, . ) .. . ,0W j...A...- * . ;,J!. a +>
SURVEILLANCE REQUIREMENTS (Continued) rrior tu iniud vperation.abovt S% RATED-THERMALTOWER after each 7 ( fueQadingr-by consideration of'tFfrctors-of-e, below...with the
- CEA groups at the Transient Insertion Limits of Specification'3 2:3.6j ce s 9 }. When in MODE 3.e6 4,^at least once per 24 hours by consideration of at least Me following factors:
- 1. React 6r' Coolant System boron concentration,
- 2. CEA position, .
- 3. Reactor Coolant System average tempdrature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted ( values to demonstrate agreement within + 1.0% delta k/k at least once per
\ 31 Effective Full Power Days (EFPD). This comparison shall consider at least i
' those factors stated in Specification 4.1.1.1.1, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.
LL%Q$. O (Q O
\
S
,PALOVERDE-UNIT /I
~
2/' -let '
.... .,..M -
. . .a.aa ~. '. . J ' x
REACTIVITY CONTROL SYSTEMS gy., A f yy' CcM C M H M W C SHUTDOWN MARGIN - y ,gLESS-THAN 4R-EQUAL--TO-210*F-LIMITING CONDITION FOR OPERATION
' J.1.1. 2 The SHUTDOWN MARGIN shall be greater than or equal to 4.0% '
delta k/k. , AP TCABILITY: MODE S gg AWSEd Mf# ACTION: With the SHUTOOWN MARGIN less than 4.0% delta k/k, immediately initiate and continue boration.at greater than or equal to 26 gpm to the reactor coolant system of a solution containing greater than or equal to 4000 ppm boron or equivalent until the' required SHUTOOWN MARGIN is restored. Ns lSURVEILLANCEREQUIREMENTS \N '
' x'
( 4.1.1.2 The SHUTOOWN MARGIN shall,be determined to be greater than or equal to 4.0% delta k/k: 7 N
~.
- a. Within 1 hour after de' tection of an' inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable.
If the inoperable-CEA is immovable as a result of excessive friction
},
or mechanical interference or known to be untrippable, the above required SHUT 00WN MARGIN shall be increased by an amount at least equal to the/ withdrawn worth of the immovable or untrippable CEA(s). N l b. At least once per 24 hours by consideration of the,following c factorst - N
,/
- 1. Reactor Coolant System boron concentration,
'N -
- 2. f CEA position, \
- 3. Reactor Coolant System average temperature, N
I 4.
\
Fuel burnup based on gross thermal energy generation, \, i
#/ 5. Xenon concentration, and j 6. Samarium concentration. i i
\
\ \
\
I PALO VERDE - UNIT 2 3/4 1,3'Z
AEAC7;vt7y c NTROL SYS Yus (O SHU7tCW waRGIN - Kn_1 - ANY CEA WITHORAWN LIMITTNG CONDITICN FOR OPERATION 3.1.1.2 ' a. The SHUTOOWN Figure 3.1-1A, andMARGIN shall be greater than or equal to that shown in b. For Tcold less than or equal to 500 oF, K N .] shall be less than 0.99. APPLICABILITY: MODES 1, 2*, 3*, 4*, partially withdrawn. and 5* with any full-length CEA fully or ACTION: a. With the SHUTDOWN MARGIN less than that in Figure 3.1-1 A, imediately
- initiate and continue boration at greater than or equal to 26 gpm to the reactor coolant system of a solution containing greater than or eoual to 4000 MARGIN ppm ofand is restored, boron or eouivalent until the required SHUTOOWN
- b. With Teola less than or equal to 500 oF and XN .' greater than or equal to 0.09, Imediately vary CEA positions and/or <
nitiate and continue boration at greater than or eoual to 26 gpm to the reactor coolant system of a solution containing greater than or eoual to 4000 ppm of baron or equivalent until the reouired K N-1 is restored. SURv!!LONCE RECUIRf9?NT3 4.1.1.2.1 With any full-length CEA fully or partially withdrawn, the SHUTOOWN MARG shall be detemined to be greater than or equal to that in Figure 3.1.1 A: a. Witnin 1 hour after cataction of an inoceracia CIA (s) and at least once per 12 tours nereafter wnile tne CIA (s) is incoeracle. If the ino:eranle CIA is immovaele as a result of excessive frietten or mecnanical interference or known to be untrip; ante, the amove required SHUTCC'dN MARGIN snall be increased my an amount at least equal to the witacrawn worth of the im.ovaole or untrip;a:la CIA (s). See Special Test Exceptions 3.10.1 and 3.10.9. PALO VERDE - UNIT / > 3/41-2
1 l
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0 100 200 300 400 500 600
. COLD LEG TDtPERATURE (*F)
FIGURE 3.1 - 1A SHUTDOWN MARGIN VERSUS COLD LEG TDtPERATURE r l f PALO VERDE - UNIT 2 3/4 1-2a .
, - -- ~ ~ m ~ ,, , - I. .,, - - _ _ _
- w. . , , , .
E., ( REAC7?'II**/ CC'17:CL TVST~wS -- SURV!!'.UNCE RECU tEMEN71 (C:nti uec) - - - . --
- h. .
. . . .... When in McCE 1 or McCE 2 with X u getatar snan er ecual t: 1.0, at least onca :er 12 neurs my verifying :nat CIA gr:u: wi uerawal is witain na Transient Insertico Limits of Specifica 1ca 3.1.3.5.
. c. When in McCE 2 wita X less taan 1.C. wit..in a neurs :rior .:
acnievingreact:rcrilyl:211ty by verifying sna ue :.ecietac , critical CIA ; sition is witnin :ne limits of Specificati:n 3.1.3.5. d.. . Price :: initia*, cetratten a:cve 5% RATED THERMAt. FCWER aftar eaca fuel Icacing, CEA ty c:nsiceratten gr u:s at tne Transian: of :ne facters of e. telcw, wiu ne Inserti:n Limits of S:ecificati:n 2.1.3.5. e. When in MC05 3, 4, or 5,at least once oer 24 hours by censideration of at leas :ne felicwing facu rs:
- 1. Reac::r C:: Tan: Systa= =ce:n :=cantra-1:n,
- 2. CIA ; sitten,
- 3. React:r C: elan: Systa: average tam:erature, 4 Fuel turnu: =asac :n gr:ss uer=al energy generati:n,
- 5. Xenen ::ncantratten, anc
- 6. Samarium c:ncan:.atien.
4.1.1.2.2 When in l'OCE 3, 4, or 5, with any full-length CEA fully or partially withdrawn, and Tcoid less than or equal to 500 C F, X .1 shall be N determined to be less than 0.99 at least once per 24 hours by consideration of at least the follcwinn factors: 1. 2. React:rC:clantSysta=::ren::ndantra.icn, CEA ::siti:n...
- 3. React.:e C:: tan: Systa: average tam:eraturt .
4
- 5. Fuel turnus :asec n gr:ss sner:al energy g. enerati:n, Xenen ::nenntratien, anc '
- 6. Sa:arium ::nesn:rati:n.
- 4.1.1.2.3 values :: The everall c:ra reactivity balance snail be c::: arac :: =recic:ac 31 Effac f ve demonstra:a Full Fewer agreement Days (!?70).wi uin - 1.C% celta k/k at least enca :er these fact:rs sta:ac in Sceci fica:f en a.1.1.2. la or 4.1.1.2.2. The pr reactivi:y values sna11 te acjustac (cereali:sc) t:
c:re c:nciticas prior :: c:rres: enc :: :ne actual loacing. ex:cecing a fuel turnu; cf 50 !?F0 aftar ea:n fuel pat.3 y!RCE - UNIT / *K 3/4 1-3' g *
. REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:
- a. A gravity feed flow path from either the refueling water tank or the spent fuel pool through CH-536 (RWT Gravity Feed Isolation Valve) and a charging pump to the Reactor Coolant System,
- b. A gravity feed flow path from the refueling water tank through CH-327 (RWT Gravity Feed / Safety Injection System Isolation Valve) and a charging pump to the Reactor Coolant System,
- c. A flow path from either the refueling water tank or the spent fuel pool through CH-164 (Boric Acid Filter Bypass Valve), utilizing gravity feed and a charging pump to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and-borated-to-a-SHUTDOWN-llARGIN-equivalent--to-at-4 east 6Ldelta-k/k-- (r" at 1109F-within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2.1 At least two of the above required flow paths shall be demonstrated OPERABLE: ,
- a. At least once per 31 days by verifying that erch valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. At least once per 18 months when the Reactor Coolant System is at normal operating pressure by verifying that the flow path required by Specification 3.1.2.2 delivers at least 26 gpm for 1 charging pump and 68 gpm for two charging pumps to the Reactor Coolant System.
4.1.2.2.2 The provisions of Specf fication 4.0.4 are not applicable for entry into Mode 3 or Mode 4 to perform the surveillance testing of Specification 4.1.2.2.b provided the testing is performed within 24 hours af ter achieving normal operating pressure in the reactor coolant system. PALO VERDE - UNIT 2 3/4 1-8 g .
CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: ,. With only one charging pump OPERABLE, restore at least two charging pu OPERABLE SHUTDOWN-MARGIN statusequivalent-to-at-least.6%. within 72 hours or bedelta. in atk/kleast HOT at.210*F STAN within the 6 hours; restore at least two charging pumps to OPERABLE status within the 7 days or be in COLD SHUTDOWN within the next 30 hours. k SURVEILLANCE REOUIREMENTS 4.1.2.4 by Specification 4.0.5.No additional Surveillance Requirements other than those req 9 0 PALO VERDE - UNIT 2 3/4 1-10,
80 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 Each of the following borated water sources shall be OPERABLE:
- a. The spent fuel pool with:
1. A minimum borated water volume as specified in Figure 3.1-2, and 2. A boron concentration of betw<.en 4000 ppm and 4400 ppm boron, and
- 3. A solution temperature between 60*F and 180'F.
- b. The refueling water tank with:
1. A minimum contained borated water volume as specified in Figure 3.1-2, and 2. A boron concentration of between 4000 and 4400 ppm of boron, and
- 3. A solution temperature between 60*F and 120*F.
. APPLICABILITY: MODES 1, 2,* 3,* and 48 ACTION:
a. With the above required spent fuel pool inoperable, restore the pool to OPERABLE status within 72 hours or be in at least HOT STANOBY !g te e SHBIOO*W% MARGIN e % elent-- k within i.o o theeit.
;=a4L next L/k e6 210af,'k hours amHorat;Q" restore the abov o to OPERABLE status within the next 7 days or be in COLD SHUT 00WN within the next 30 hours, b.
With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.6 OPERABLE: Each of the above required borated water sources shall be demonstrated
- a. At least once per 7 days by:
1. Verifying the boron concentration in the water, and 2. Verifying the contained borated water volume of the water source. . b. At least once por 24 hours by verifying the refueling water tank temperature 120"F range. when the outside air temperature is outside the 60*F to c. At least once per 24 hours by verifying the spent fuel pool temperature when irradiated fuel is present in the pool. See Special Test Exception 3.40.7. PALO VERDE - UNIT 2 3/4 1-13 I . .
l 1 TABLE 3.3-1(Continued) REACTOR PROTECTIVE INSTRUMENTATION TABLE NOTATIONS l
*With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.
#The provisions of Specification 3.0.4 are not applicable.
(a) Tripmaybemanuaihbypassedabove104%ofRATEDTHERMALPOWER; bypass shall be automat call equal to 10 4% of RATED THER L POWER. removed when THERMAL POWER is less than or l (b) Trip may be manually by)assed below 400 psia; bypass shall be l ' automatically removed wienever pressurizer pressure is greater than or
' equal to 500 psia.
to% (c) Trip may be manually bypassed below 1X of RATED THERMAL POWER bypass shall be automatically removed when THERMAL POWER is gr; eater than or equal to 1% of RATED THERMAL POWER. Ia n (d) Trip ma 3.10.3.y be bypassed during testing pursuant to Special Test Exception (e) See Special Test Exception 3.10.2. (f) There are four channels, each of which is comprised of one of the four reactor trip breakers arranged in a selective two-out-of-four configuration (i.e. , o,ne-out-of-two taken twice). l l ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement restore the inoperable channel to OPERABLE status within 48 hours, or be in at least HOT STANDBY within the next 6 hours and/or open the protective system trip breakers. ACTION 2 - With the number of channels OPERABLE one less than the Total Number of Channels STARTUP and/or POWER'0PERATION may continue provided the inoper,able channel is placed in the bypassed or tripped condition within 1 hour. If the inoperabic channel is ' ' bypassed, the desirability of maintaining this channel in the bypassed condition shall be reviewed in accordance with S accification 6.5.1.6.g. The channel shall be returned to 0)ERABLE status no later than during the next COLD SHUT 00WN. L PALO VERDE - UNIT 2 . 3/4 3-5, I ' l _- . _ J - _ __ - -
3/4.10 SPECIAL TEST EXCEPTIONS
\ 3/4.10.1 SHUTDOWN MARGIN ,4ap go_, &d ANR/77 ecW6 LIMITING CONDITION FOR OPERATION fMdKp/ 1 3.10.1 The SHUTDOWN MARGINarequirementofSpecification3.1.1.fmaybe suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s), or the reactor is subcritical by at least the reactivity equivalent of the highest CEA worth.
APPLICABILITY: MODES 2, 3* and 4*#. ACTION: a. With any full-length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immedi-ately initiate and continue boration at greater than or equal to 26 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN / required by Specification 3.1.1.J restored. puo ,<y ,
- b. With all full-length CEAs fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 26 gpm of a solution containing greater than or equal to 4000 ppm baron or its equivalent I(k until the SHUTOOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length and part-length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. 4.10.1.3 When in MODE 3 or NODE 4, the reactor shall be determined to be subcritical by at least the reactivity equivalent of the highest estimated CEA ' worth or *.he reactivity equivalent of the highest estimated CEA worth is avail-able for trip insertion from OPERABLE CEAs at least once per 2 hours by con-sideration of at least the following factors:
- a. Reactor Coolant System boron concentration,.
- b. CEA position,
- c. Reactor Coolant System average temperature,
- d. Fuel burnup based on gross thermal ener0y gene' ration,
- c. Xenon concentration, and
- f. Samarium concentration.
A Operation in M30E 3 and MODE 4 shall be limited to 6 consecutivo hours. Limited to lo,e power PHYSICS TESTING at the 320*F plateau. PALO VERDE - UNIT 2 3/4 10'1 l . -
314.10 SPECIAL TEST EXCEPTIONS 3/4.10.9 SHUTUCW hARGIN AND XN-1. CEDMS TESTIN l,IMITING CCNOITION FOR OPERATION 3.10.9 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 a SHUTDOWN MARGIN and K -1 requirements of Specification 3. N for pre-startup drive tests mechanism system to demonstrate provided: the OPERABILITY of the contr
- a. No more than one CEA is withdrawn at any time.
- b. No CEA is' withdrawn more than 7 inches.
- c. the Thestart KN-1 of requirement testing, of Specification 3.1.1.2 is met prior to
- d. suspended All other operations involving positive reactivity changes are during the testing.
APPLICABILITY: MODES 4 and 5. ACTION: With any of the above requirements not met, suspend testing and comply with the requirements of Specification 3.1.1.1 or 3.1.1.2, as app
$URVETLLANCE RECU?REMENTS 4.10.9 within one hour prior to the start of testing, and at lea during testing.
4 PALOVER0E-UNIT /% , 3/4 10 9 A
3/4.1 REACTIVITY CONTROL SYSTEMS BASES Y 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1 2
'SHUTD0WN MARGIN /tldD h A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all ope'r'ating conditions (2) the reactivity transients /,/
associated with postulated accident conditions are controllable within accept-able limits assuming the insertion of the regulating CEAs are within the limits of Specification 3.1.3.6, and (3) the reactor will be maintained sufficiently condition. subcritical to preclude inadvertent criticality in the shutdown
$ff .9E/4SG Amf ~;Y W&MG ,.. /
- SHUT 00WN MARGIN requirements vary throughdu%' t tore life as a function of fuel depletion, RCS boron concentration, and RCS T coid. The most restrictive
; condition occurs at E0L, with T cold at no load. operating temperature, and is associated with a postulated steam line brea'k accident and resulting uncon-trolled RCS cooldown. In the analysis.of this accident, a minimum SHUTDOWN MARGIN of 6.0% delta k/k is required'to control the reactivity transient.
Accordingly, the SHUT 00WN MARGIN' requirement is based upon this limiting [ condition and is consistent.with the criteria used to establish the power dependent CEA insertion. limits and with the assumptions used in the FSAR ( Safety Analysis. ' i With i j[ess than or equal to 210*F, the reactivity transients resulting I , from uncontrolled RCS cooldown are minimal and a 4% Ak/k SHUTDOWN MARGIN
.requipement is set to ensure that reactivity transients resulting from an
@nadvertentsingleCEAwithdrawaleventareminimal.
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (MTC) 4 The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle. The surveillance requirements for - measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that ( the measured MTC value is within its limit provides assurances that the coef-l j
\ ficient will be maintained within acceptable values throughout each fuel cycle. f
-f
* &t PALO VERDE - UNIT 2 8 3/4 1,1
; I .
hN k ( 3/4.1 REACTIV!TY CtN770L SY5795 '
~_..-
Basis ' _ _ . . _ a 3/4.1.1 !C.7ATION CtN770L -- 3/1.1.1.1 and 3/1.1.1.2 SFUT*CW mGIN ANO Ky,,)
- - .- ~ ~~ - $
' suberitical following a design basis accident or anticipated occurrence. The function ofN K -1 is to maintain sufficient suberiticality to preclude inadvertent criticality following ejection of a single control element assembly (CEA). During coeratien in MODES 1 and 2, with keff greater than or equal to 1.0, the transient insertion limits of Specification 3.1.3.6 ensure that sufficient SHUTCOWN MARGIN is available. SHUTCOWN MARGIN is the amount by which the core is subcritical, or malfunction resulting in the highest worth CEA failing t N measure in the highest of the core's worth reactivity, inserted CEA beingconsidering ejected. a single malfunction resulting SHUT 00WN MARGIN requirements vary throughout the core life as a function (T of fuel depletion and reactor coalant system (RCS) cold leg temoeratu cold). The most restrictive condition occurs at EOL, with T,,,jg at no-load operacing temperature,and is associated with a postulated steld sine break accident and the resulting uncontrolled RCS cooldewn. In the analysis of this accident, the specified SHUTCOWN MARGIN is required to control the reac satisfied. As (initial) Teoldtransient and ensure that the fuel perfom resulting reactivity transient are less severe and, therefore, the re SHUT deboration 00WN eventMARGIN also decreases. beccmes limiting wi ofBelow Ten)a about 210 oF the inadvertent requirements. Below 210 0F, the spec.th respect to the SHUTCOWN MARGIN ified SHUTCOWN MARGIN ensures that sufficient time for operator actions exists between the initial indication of the deboration and the totsl loss of shutdeun margin. Accordingly, with at leas ' one CEA uponpartially or fullyconditions. withdrawn, the SHUTCOMN MARGIN requirements are based these limiting , MARGIN that are not limiting with respect to the Specifica single CEA withdrawal and startup of an inactive reactor coolant pump. Kg.1 requirements vary with the amount of positive reactivity that would be introduced assuming the CEA with the highest inserted worth ejects fr the core. In the analysis of the CEA ejection event, the KN .1 requirement considering power redistributien effects. Above Teold of 500 reactivity feedback is sufficient to preclude the need for a specific Kg. requirement. With all CEAs fully inserted, Kn.1 and SHUTCCWN MARGIN requ are equivalent in terms of minimum acceptable core boren concentration. PALO V!X:5 - UN!7 J 1 6 3/4 1-1
3/4.1 RE.sCTIVITI CON 770L SYSTIus
!ASES 3/A.1.1 ! ORATION CONTROL (cont.) '
3/4.1.1.1 sed 3/4.1.1.2 SMUTCC'a .uA: GIN AND Kn 1 (ccra.L___. .___._ _ _. SHUTDOWN fiARGIN N-1 are: or KOther technical specifications that reference the Sp CONTROL ASSEMBLIES, 3/4.9.1, REFUELING OPERATIONS-3/4.10.1, SHUTCOWN MARGINN AND K CEA WORTH TESTS, ,
, SHUTCOWN and 3/a.10.9 MARGIN AND K CECMS TESTING.
N 3/1.1.1.3 :
.ve. r .:. 4 7. ., i .:.9..: ::.A . "..:. :. r. .. .:.:: ... . .1,
. ru-.)
The Ifmitniens en to ensure sna; :ne asse:::cern:r :ac:aratura c:affician:::icns nsien analysis usac in :na ac:i:a remain valic thr:ugn eac: fuel cycle. The surveillanca .scuiremen.s fer measurement of ine MTC curing eac: fuel cycle art acecua:a :: c:nfi m ::a MTC value sinca Mis c:afficient enangas s:cwly cue ;rinci;aily :: :ne recucti:n in RC5 acren c:ncan:rnien ass: cia .ac with fuel The burnu:. the measured MTC value is witnin c
- s tini; pr:vicas assuranca :nfir a-ica ::a-ficient cycla. will be maintainac witnin ac:a::n le values :nr:ugn us eacn :na.fuel :ne ::af-3/4.1.1.4 MINI."et Ti.M:!RArdRE .:02 Cx!TicAL!rt . _ _ . . _
This specificnien ensures na: i is recuired to ensurethe 1 Reacter C: clan: Systam c:lc leg taccaratura less This Ifatta-ion } analy:sd ta=;ernura ra(n) :na mocara :r tam:erature c: efficient is wi nin ita nor=al ocerating range, ge, (2) :ne pretac.ive instru=anta:f =n is wi:nin i:s analysis. and (3) to ensure c:nsistancy wita :ne FIAR safa y I l PALO VE:.CE - UNIT $ 3 3/41-la . i .
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY - This specification ensures that the reactor will not be made critical with7 the Reactor Coolant System cold leg temperature less than 552 F. is required to ensure (1) the moderator temperature coefficient isThis limitation within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range,-and (3) to ensure consistency with the FSAR safety analysis. _ 3/4.1.2 BORATION SYSTEMS ) s f g g_b The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) an emergency power supply from OPERABLE diesel generators, and (5) the volume control tank (VCT) outlet valve CH-UV-501, , capable of isolating the VCT from the charging pump suction line. The nominal capacity of each charging pump is 44 gpm at its discharge. Up to 16 gpm of this may be diverted to the volume control tank via the RCP control bleedoff. Instrument yielding the 26inaccuracies gpm value. and pump performance uncertainties are limited to 2 gpm With the RCS temperature above 210 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability able. in the event an assumed failure renders one of the systems inoper-Allowable out-of-service periods ensure that minor compcnent repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. nsesTha Srs22M is cWH8uf of Moi 4Dwt Bona?c/JcTMw a suT'70 A hnration Capabi.lity nf.eithar4ystem..iS sufficient to provide -a-SHUTDOWN MARGIN from expect decay and cooldown to 210 F eratinn-ronditions-of 4% delta k/k af ter xenon The maximum expected boration capability require-ment occurs at E0L from full power equilibrium xenon conditions and requires 23,800 gallons or the spent of pool. fuel 4000 ppm borated water from either the refueling water tank gst (nw @ , With the RCS temperature below 210 F one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable. The restrictions of one and only one operable charging pump whenever reactor coolant level is below the bottom of the pressur-izer is based on the assumptions used in the analysis of the boron dilution 1 Mh9 - A-The-boron--capabi4i-ty-required below-210 F-is based upon providing a 4% delta-k/k -SHUTOOWN-MARGIN-af-ter- xenon-decay and cooldown from 210 F to 120 F. This condition requires 9,700 gallons of 4000 ppm borated water from either the refueling water tank or the spent fuel pool. PALO VERDE - UNIT 2 8 3/4 1-2
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3/4.10 SPECIAL TEST EXCEPTIONS BASES lf( 3/4.10.6 SAFETY INJECTION TANKS Thissystem injection specialcheck test exception valves. The permits testing the low pressure safety pressure in the injection header must be reduced below the head through the check valves: of the low pressure injection pump in order to get flow The safety injection tank (SIT) isolation valve must be closed in order to accomplish this. The SII isolation valve is still capable of automatic operation in the event of an SIAS; therefore, system capability should not be affected. 3/4.10.7 SPENT FUEL POOL LEVEL This special test exception permits loading of the initial core with the spent fuel pool dry. 3/4.10.8 SAFETY INJECTION TANK PRESSURE This special test exception allows the performance of PHYSICS TESTS at low pressure / low temperature (600 psig, 320 F) conditions which are required to verify the low temperature physics predictions and to ensure the adequacy of design codes for reduced temperature conditions.
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H. Supporting Analyses for the Technical Specification Amendment Request:
- 1. STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT - MODE 3 OPERATION: TC <500*F 1.1 Identification of Event and Causes Refer to CESSAR Section 15.1.4, except that steam line break events during Mode 3 operation for reactor cold leg temperature (Teold) less than 500*F are analyzed to demonstrate the adequacy of the shutdown margin as specified by Technical Specifications 3.1.1.1 and 3.1.1.2, to prevent degradation in fuel performance as a result of post trip return to power. The results show that the shutdown margin is sufficiently large to prevent a post trip return to power. The steam line breaks presented are:
A. A large steam line break inside containment during Mode 3 operation with concurrent loss of offsite power in combination with a single failure and 3 technical specification shutdown margin. 4 B. A large steam line break inside containment during Mode 3 operation with offsite power available in combination with a single failure and technical specification shutdown margin. t For T > 500*F, the shutdown margin specified by Technical Specification eold j 3.1.1.2 is 6% A p . Below 500*F, the shutdown margin decreases linearly with temperature. Here, cold leg temperatures below 500*F are considered. Hot zero power steam line breaks above 500*F using technical specification shutdown margin are found in PVNGS FSAR Section 15.1.5. The requirements of Technical Specification 3.1.1.1 are less limiting for steam line breaks than are those of Technical Specification 3.1.1.2. The largest possible steam line break size is the double ended rupture of a steam line upstream of the main steam isolation valve (MSIV). In the PVNGS design, an integral flow restrictor exists in each steam generator outlet nozzle. The largest effective steam blowdown area for each steam line, which is limited by the flow restrictor throat area, is 1.28 square feet.
These two cases are analyzed for end of equilibrium core, self-generated plutonium recycle (SGR) conditions. PVNGS specific minimum safety injection flow rates, feedwater isolation valve closure time, and steam generator differential pressure ( A P) isolation (lockout) setpoint are employed. 1.2 Sequence of Events and System Operation Steam line breaks are characterized as cooldown events due to the increased steam flow rate, which causes excessive energy removal from the steam generators and the reactor coolant system (RCS). This results in a decrease in reactor coolant temperatures and in RCS and steam generator pressures. The cooldown causes an increase in core reactivity due to the negative moderator and Doppler reactivity coefficients. i Mode 3 steam line breaks are initiated from a suberitical reactivity condition. Detection of the cooldown is accomplished by the pressurizer and steam generator low pressure alarms, by the high reactor power alarm and by the low steam generator water level alarm. Reactor trip is provided by one of two available reactor trip signals. These are the low steam generator pressure and the high logarithmic power level trips. For a steam line break that occurs with a concurrent loss of offsite power, termination of feedwater to both steam generators and coastdown of the reactor coolant pumps are assumed to be initiated simultaneously. In general, the depressurization of the affected steam generator results in actuation of a main steam isolation signal (MSIS). This closes the MSIVs, isolating the unaffected steam generator from blowdown, and closes the main feedwater isolation valves (MFIV), terminating main feedwater flow to both steam generators. After the reduction of steam flow that occurs following MSIV closure, the level in the intact steam generator falls below the auxiliary feedwater actuation signal (AFAS) setpoint. The resulting AFAS causes auxiliary feedwater (AFW) flow to be initiated to both steam generators. If the differential pressure between the two steam generators exceeds the setpoint, the AFW logic isolates flow to the affected steam generator and
.l diverts the flow from both AW pumps to the intact steam generator. The j . pressurizer pressure may decrease to the point where a safety injection actuation signal (SIAS) is initiated. The isolation of the unaffected steam generator and subsequent emptying of the affected steam generator terminate the cooldown. The introduction of safety injection boron upon SIAS causes core reactivity to decrease. The operator, via the appropriate emergency procedures, may initiate plant cooldown by manual control of the atmospheric l steam dump valves, or, in the event that offsite power is available, by using the unaffected steam generator and the turbine bypass valves, any time af ter the affected steam generator empties. The analyses presented herein conservatively assume operator action is delayed until 30 minutes after event initiation. The plant is then cooled to 350*F and 400 psia, at which point shutdown cooling is initiated. 1- ! A parametric study of single failures (see Appendix 15C of CESSAR) that would have an adverse impact on the SLB event has determined that the-failure of one of the high pressure safety injection (HPSI) pumps to start following SIAS has I the most adverae effect for those cases that result in generation of SIAS. For the two cases presented here, there is no SIAS actuation for the duration of the transient (500 seconds). For these events the most adverse effect is caused by the failure of a MSIV on one of the steam lines from the intact generator to close following MSIS. Consequently, for these cases steam is assumed to continue to be released from the intact steam generator at 1.5% of the design steam rate. This open flow path is represented by an effective flow area for steam blowdown from the intact steam generator of 0.034 square feet. 1.3 Analysis of Effects and Consequences A. Mathematical Models ! The mathematical models and data transfer between codes used in the SLB I analysis are presented in PVNGS FSAR Appendix 15.C.
i B. Input Parameters and Initial Conditions The initial conditions assumed in the analysis of the NSSS response to ] Cases 1 and 2 are presented in Table 1-1. The initial K,gf of 0.99 is q_ the highest value allowed by technical specifications for Mode 3 and 4 leaves the least pretrip margin to criticality. There is no effect on the post trip margin to criticality. Above core inlet temperatures of 500*F, the shutdown margin is 6% Ap.. Below 500*F the required shutdown margin decreases linearly with temperature. The initial core 1 inlet temperature of 450*F was selected to demonstrate the adequacy of the shutdown margin in the temperature range where its magnitude is decreasing. This is a representative cold leg temperature. Analysis at other initial cold leg temperatures below 500*F will produce results and parameter trends similar to those presented here. Initially two reactor coolant pumps are assumed to be operating, as allowed in Mode 3. The initial pressurizer pressure of 830 psia falls within the range of normal Mode 3 operating procedures. The SIAS setpoint is set at 430 psia, 400 psi below the initial pressurizer pressure, the maximum offset allowed by technical specifications. This and the high initial pressurizer water volume have the effect of delaying SIAS actuation since SIAS generally 4 occurs after the pressurizer empties. The technical specification shutdown margin at 450*F is 5.1% Ap . Since the reactor is 1% suberitical initially, a CEA worth at trip of 4.1% Ap is assumed. The moderator and Doppler reactivity coefficients corresponding to the end of equilibrium cycle, self-generated plutonium recycle (SGR) are employed. For the purpose of conservatism the moderator reactivity coefficients correspond to the condition of no initial boron in the core. C. Results j Case 1: Large Steam Line Break During Mode 3 Operation with
! Concurrent Loss of Offsite Power (SLBM3 LOP) 4 7
i ^ 1 a
. _ - . _ , . , . , , - .,._..___.,m.__. _ , . _ _ _ _ -.___,__.__,r...,...__.___, . _ , , . . . . . , _ - - . , _ , . _ - . , , . . _
1 l l The dynamic behavior of the salient NSSS parameters following the SLBM3 LOP. is presented in Figures 1-1 through 1-16. Table 1-2 summarizes _ the major events, times, and results for this transient. 4 1 Concurrent with the steam line break, a loss of offsite power occurs. At ' this time an actuation signal for the emergency diesel generators is initiated. Also at this time, the CEDM coils are assumed to lose power and, after a 0.34 second coil decay delay, the CEAs begin to drop into the core. At 21.3 seconds the steam generator pressure falls below the j main steam isolation signal (MSIS) setpoint of 223 psia. This results in the generation of MSIS at 22.3 seconds, which initiates closure of the j MSIVs and MFIVs. The MSIVs close by 26.9 seconds. The MFIVs close by 31.9 seconds. i During the first 500 seconds of the transient the pressurizer has not yet i emptied and pressurizer pressure remains above the SIAS setpoint of 430 psia. Hence no safety injection flow and no boron reaches the RCS during this time. AFAS is assumed to be actuated soon after the MSIVs close. Auxiliary i feedwater is assumed to enter the steam generators af ter the level falls below the 80% high level setpoint; i.e., at 118 seconds. The pressure 1 difference between the two steam generators remains below the analysis setpoint of 325 paid during the transient. Hence there is no automatic isolation of auxiliary feedwater to the affected steam generator. At 500 seconds the transient reactivity is -2.1%, which indicates there 1s still a significant margin to recriticality. This margin will l ! continue to decrease as the affected steam generator continues to blow j down and the RCS continues to cool. After the pressurizer empties the RCS pressure is expected to fall more rapidly resulting in SIAS and j subsequent inflow of boron into the RCS. Alternately SIAS may be manually actuated by the operator. In either case, after the inflow of boron into the RCS, the margin to recriticality is expected to increase. v . _ .___._._-.m._.__,_,_.___-,--. _ _ . , , , _ _ , . , _ . . . _ , . - - - . - , . ._.--.,,__. - .___ _,
Eventually, the affected steam generator is expected to blow down to atmospheric pressure. This would terminate further RCS cooldown. Even assuming the limiting case, where the affected steam generator has j depressurized to atmospheric pressure and no safety injection boron has reached the RCS, the core will remain suberitical with a margin to criticality of no less than -0.4% Ap. The discontinuity seen in some of the parameter plots at about 470 seconds (e.g., Figures 1-6 and 1-8) is due to safety injection tank (SIT) flow into the RCS for a short period of time. A SIT injection gas cover pressure of 608 psia was used in the analysis. The effect of this is small since no credit was taken for the SIT boron in the analysis. The minimum DNBR remains above 10 during this transient. At a maximum of 30 minutes, the operator, via the appropriate emergency procedure, initiates plant cooldown by the manual control of the atmospheric dump valves. Shutdown cooling is initiated when the RCS reaches shutdown cooling entry conditions. Case 2: Large Steam Line Break During Mode 3 Operation with Offsite Power Available (SLBM3) The dynamic behavior of the salient NSSS parameters following the SLBM3 is presented in Figures 1-17 through 1-32. Table 1-3 summarizes the major event, times, and results for this transient. At 24.3 seconds af ter the initiation of the steam line break, the steam generator pressure drops below the low steam generator pressure trip and MSIS setpoint of 223 psia. The reactor trip breakers open at 25.45 seconds. After a 0.34 second coil delay, the CEAs begin to drop into the core at 25.8 seconds. The MSIS initiates closure of the MSIVs and MFIVs. The MSIVs close by 29.9 seconds. The MFIVs close by 34.9 seconds. During the first 500 seconds of the transient, as in Case 1, the pressurizer has not yet emptied and pressurizer pressure remains above the SIAS setpoint of 430 psia. Hence no safety injection flow and no safety injection boron reaches the RCS during this time. AFAS is assumed to be actuated soon af ter the MSIVs close. Auxiliary feedwater is assumed to enter the steam generators after the level falls below the 80% high level setpoint, i.e., at 115 seconds. The pressure difference between the two steam generators stays below the analysis setpoint of 325 paid during the transient. Hence there is no automatic isolation of auxiliary feedwater to the affected steam generator. At 500 seconds the transient reactivity is -1.9%, which indicates there is still a significant margin to criticality. This margin will continue to decrease as the affected steam generator continues to blow down and the RCS continues to cool. After the pressurizer empties the RCS pressure is expected to fall more rapidly resulting in SIAS and subsequent inflow of boron into the RCS. Alternately SIAS may be manually actuated by the operator. In either case, after the inflow of boron into the RCS, the margin to recriticality is expected to increase. Eventually, the affected steam generator is expected to blow down to atmospheric pressure (T=212*F). This would terminate further RCS cooldown. Even assuming the limiting case, where the affected steam generator has depressurized to atmospheric pressure and no safety injection boron has reached the RCS, the core will remain suberitical with a margin to criticality of no less than -0.4% Ap. The discontinuity seen in some of the parameter plots at about 390 and 450 seconds (e.g., Figures 1-22 and 1-24) is due to safety injection tank (SIT) flow into the RCS for a short period of time. A SIT injection gas cover pressure of 608 psia was used in the analysis. The effect of this is small since no credit was taken for the SIT boron in the analysis. The minimum DNBR remains above 10 during the transient. At a maximum of 30 minutes, the operator, via the appropriate emergency procedure, initiates plant cooldown. Shutdown cooling is initiated when the RCS reaches shutdown cooling entry conditions. 1,4 Conclusion For the large steam line break during Mode 3 operation for reactor cold leg temperatures less than 500*F with or without a loss of offsite power, and in combination with a single failure the shutdown margin is sufficient to prevent a post trip return to power.
TABLE 1-1 ASSUMPTIONS AND INITIAL CONDITIONS FOR LARGE STEAM LINE BREAKS DURING MODE 3 GPERATION WITH AND WITHOUT CONCURRENT LOSS OF OFFSITE POWER (SLBM3 LOP & SLBM3) Parameters Assumed Value Initial Reactivity 0.99 Initial Core Inlet Coolant Temperature, F 450 Initial Core Mass Flow Rate, 106 lbm/hr (2 RCPs) 91.1 Initial Pressurizer Pressure, psia 830 Initial Pressurizer Water Volume, ft3 1100 Doppler Coefficient Multiplier 1.15 1 Moderator Coefficient Multiplier 1.10 Axial Shape Index +.3 CEA Worth at Trip, 10-2 Ap -4.1 Initial Steam Generator Inventory, iba 311,000 Core Burnup End of Cycle Blowdown Fluid Saturated Steam ! Blowdown Area for Each Steam Line, ft 2 1.283
+ - - , . , . - , - ----..--.-------,--,,7--.-n. - . _ - - . --
.-----r - - . .-- , ,-- ~.-,-,-,. g , , , , -.
i TABLE 1-2 1 SEQUENCE OF EVENTS FOR A LARGE STEAM LINE BREAK DURING MODE 3 OPERATION WITH CONCURRENT LOSS OF 0FFSITE POWER (SLBM3 LOP) i Time (Sec) Event Setpoint or Value i 0.0 Steam line break and loss of offsite power occur. Holding coils lose power. -- 21.3 Steam generator pressure reaches main 4 steam isolation signal (MSIS) analysis
! setpoint, psia 230 22.3 MSIS generated --
l 26.9 MSIVs completely closed -- 31.9 MFIVs completely closed -- 500 Transient reactivity,10-2 A p -2.1 2
> 500 Pressurizer empties --
i j > 500 Pressurizer pressure reaches safety injection actuation signal (SIAS) analysis setpoint, psia 430
> 500 SIAS generated --
> 500 Voids begin to form in reactor vessel upper head --
> 500 Safety injection flow begins --
> 500 Safety injection boron begins to reach reactor core 1800 Operator initiates cooldown --
t 1 _-- .- _ _ ,._.. .-___ _ _ , __ _ _. , . - .._,_ _ _ ., _ ._____ _ . _,_ _ . _ _ _ , _ . . _ _ . _ _ ~ _ _ _ _ _ .
i l TABLE 1-3 SEQUENCE OF EVENTS FOR A LARGE STEAM LINE BREAK DURING MODE 3 OPERATION WITH OFFSITE POWER AVAILABLE (SLBM3) Time (Sec) Event Setpoint or Value 0.0 Steam line break occurs -- 24.3 Steam generator pressure reaches main steam isolation signal (MSIS) analysis setpoint and low steam generator pressure trip setpoint,-psia 230 25.3 Low steam generator pressure trip signal and MSIS generated -- 25.45 Reactor trip breakers open -- 29.9 MSIVs completely closed -- 34.9 MFIVs completely closed -- 500 Transient reactivity,10-2 a p- _1,9
> 500 Pressurizer empties --
> 500 Pressurizer pressure reaches safety injection actuation signal (SIAS) analysis setpoint, psia 430 ,
> 500 SIAS generated --
> 500 Voids begin to form in reactor vessel
. upper head --
, > 500 Safety injection flow begins --
> 500 Safety injection boron begins to reach reactor core 1800 Operator initiates cooldown --
J
- 2. STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT - MODE 4 OPERATION 2.1 Identification of Event and Causes i
- The steam system piping failure event during Mode 4 operation is evaluated to demonstrate the adequacy of the shutdown margin, as specified by technical
; specifications, to prevent degradation in fuel performance as a result of a post trip return to power.
? The double ended rupture of a steam line upstream of the main steam isolation valve (MSIV) is considered. Reactor physics parameters (e.g., moderator and f - Doppler reactivity coefficients) for the end of equilibrium core, self generated plutonium recycle (SGR) are assumed. The moderator reactP.*ity coefficient corresponding to the condition of zero initial boron in the core
- f. is conservatively assumed. The more likely condition of boron existing in the l RCS would cause the moderator reactivity to be less negative or even positive, thus reducing the reactivity increase during a steam line break.
Extrapolation of the results from Section 1 show that the shutdown margin in Mode 4 is sufficiently large to prevent a post trip return to power. l l 2.2 Sequence of Events and Systems Operation I' Steam line breaks result in excessive cooldown of the reactor coolant which 3 causes an increase in core reactivity due to the negative moderator and Doppler reactivity coefficients. t i Mode 4 steam line breaks are initiated from a suberitical reactivity condition. The initial cold leg temperature may range from 210'F to 350*F. The sequence of events during a steam line break event will differ depending - on the initial cold leg temperature and pressurizer pressure. I Detection of the cooldown is accomplished by the pressurizer and steam generator low pressure alarms and by the low steam generator water level alarm. I 4
,n,- ow-,- ,. ,- n,,--m--~g,-- - - , , ~ ,,,,~,--,,,,w-.---,, , , , - - , - ~ , - - , . ~ , - , - ----.---..,.,,,w,w-3- ,_n ,,_,,,-,,,,,--,y - + - - , -,,-,,,m_,pn-- - , - , - , , - - -
i Reactor trip is provided by one of two available reactor trip signals. These are the low steam generator pressure and the high logarithmic power level trips. The depressurization of the affected steam generator may result in actuation of a main steam isolation signal (MSIS). This closes the MSIVs, isolating the unaffected steam generator from blowdown. If the level in either steam generator falls sufficiently low, an auxiliary feedwater actuation signal (AFAS) will occur. The pressurizer pressure may decrease to
; the point where a safety injection actuation signal (SIAS) is initiated. The i ! introduction of safety injection boron upon SIAS causes core reactivity to decrease. Eventually the affected steam generator will depressurize to atmospheric pressure which will terminate the RCS cooldown.
l 2.3 Analysis of Effects and Consequences The magnitude of the core reactivity increase during a steam line break during 3 Mode 4 depends on the initial cold leg temperature. A steam line break initiated from 210*F or lower will result in negligible increases in core
- reactivity. A steam line break initiated from 350*F has the potential for
! moderate amounts of reactivity increase. Eventually, the affected steam generator will blow down to atmospheric pressure terminating further reactivity increase. If a SIAS is initiated, either automatically or by manual operator action, the core reactivity will decrease subsequent to the inflow of boron into the RCS. Even assuming the limiting case where the j
affected steam generator has depressurized to atmospheric pressure and no safety injection boron has reached the RCS, the technical specification shutdown margin will prevent a return to core criticality. 2.4 Conclusion For the large steam line break during Mode 4 operation, the shutdown margin as specified by the technical specifications is sufficient to prevent a return to f core criticality.
I
- 3. UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION I
Refer to CESSAR Section 15.4.1.
- 4. UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM MODES 2 AND 3 SUBCRuTICAL WITH 4 REACTOR COOLANT PUMPS OPERATING 4.1 Identification of Event and Causes An uncontrolled sequential withdrawal of CEAs is assumed to occur as a result of a single failure in the Control Element Drive Mechanism (CEDM), Control Element Drive Mechanism Control System (CEDMCS), reactor regulating system, or as a result of operator error. This event is analyzed to justify reduced Technical Specification Shutdown Margin requirements in suberitical modes to below that required for Modes 1 and 2.
4.2 Sequence of Events and Systems Operation The withdrawal of CEAs from suberitical conditions adds reactivity to the reactor core, causing both the core power level and the core heat flux to increase followed by corresponding increases in reactor coolant temperatures and reactor coolant system (RCS) pressure. The withdrawal of CEAs also produces a time dependent redistribution of core power. These transient variations in core thermal parameters may result in an approach to the specified acceptable fuel design limits- (SAFDL), thereby requiring the protective action of the reactor protection system (RPS). Reactivity Control The reactivity insertion rate accompanying the uncontrolled CEA withdrawal is dependent primarily upon the CEA withdrawal rate and the CEA worth since, at subcritical conditions, the normal reactor feedback mechanisms do not occur until power generation in the core is large enough to cause changes in the fuel and moderator temperatures. The reactivity insertion rate determines the rate of approach to the fuel design limits. The uncontrolled CEA withdrawal
transient from subcritical conditions is terminated by a high logarithmic power level trip or a CPC range trip. Reactor Heat Removal - Following the cooldown phase in which the steam bypass control system is used, the shutdown cooling system (SCS) is manually actuated when the RCS temperature and pressure have been reduced to 350*F and 400 psia, respectively. This system provides sufficient cooling flow to cool the RCS to cold shutdown. Primary System Integrity The RCS pressure remains below the pressurizer pressure safety valve setpoint and remains less than 110% of design pressure. The pressurizer pressure control system and the pressurizer level control system are manually opere.ted to regulate RCS pressure and coolant inventory during the cooldown phase. Secondary System Integrity The secondary system pressure increases following reactor trip and is limited by the steam generator safety valves. The atmospheric dump valves are used to cool the plant down to shutdown cooling entry conditions. The feedwater flowrate is in manual mode and is very low because it matches steam flow rates. Table 4-1 gives the sequence of events for the limiting CEA withdrawal transient from suberitical conditions identified in paragraph 4.3. 4.3 Analysis of Effects and Consequences A. Mathematical Model
'Ihe Nuclear Steam Supply System (NSSS) response to a CEA sequential withdrawal from suberitical or low power conditions was simulated using
the CESEC computer program described in PVNGS FSAR Section 15.0. The thermal margin on DNBR in the reactor core was simulated using the TORC computer program described in PVNGS FSAR Section 15.0 with the CE-1 CHF correlation described in PVNGS FSAR Qiapter 4. B. Input Parameters and Initial Conditions The initial conditions and NSSS characteristics assumed in this analysis have been determined to be the limiting conditions from which a CEA withdrawal could be initiated from suberitical modes. The range of initial conditions which were considered is limited both by Technical Specification LCO's and by the auxiliary trip functions in the CPC's. At 10 percent power, the CPC zero power bypass is automatically removed. If initial NSSS parameters or core coolant flow, temperature, pressure etc., are beyond specified values the CPC's will cause an l immediate reactor trip upon bypass removal. The 10- percent power level setpoint at which the CPC bypass is automatically removed is well below the setpoint of the high logarithmic power level trip. A trip generated at this power level would cause a decrease in fission power before the point of adding sensible heat flux is reached, thereby > precluding a challenge to the SAFDLs. Parametric studies of initial conditions which would not generate an i immediate CPC trip were performed. The initial conditions which resulted in the most adverse transient are presented in Table 4-2. All control element assemblies are initially assumed to be fully inserted and the initial K,gf is 0.91. The reactivity insertion rate is conservatively taken to be 3.23 x 10 op /second which is greater than the maximum differential worth of the highest worth CEA bank.
- . - - - . = . . - -- . . - ..- - - _ _ _. -
4 l
!. C. Results The dyncaic behavior of important NSSS parameters following a CEA withdrawal from suberitical conditions is presented in Figures 4-1 through 4-5.
The withdrawal of CEA's from suberitical conditions gradually reduces the amount by which the core is shut down. During this time, suberitical multiplication causes core power to increase. The reactor reaches critical at 278 seconds. Following this, core power rises at a rate which increases as the CEA's continue to withdraw. A reactor trip on high logarithmic power is generated before core power reaches the point of adding sensible heat. Due to the rapid rate of power increase at the time of trip generation and the effect of continued CEA withdrawal until the trip breakers open, a brief power excursion occurs past the point of adding sensible heat. 4 The CEA's begin dropping into the core at 293.7 seconds terminating the power escalation with a hot channel minimum DNBR greater than 2.0. I 4.4 Conclusions '
- The uncontrolled CEA withdrawal from a suberitical condition event meets general design criteria 25 and 20 as specified in SRP 15.4.1. These criteria j
require that the specified acceptable fuel design limits are not exceeded and that protection system action is initiated automatically. The transient terminates with a hot channel minimum DNBR greater than 1.19 and the peak fuel centerline temperature during the transient is less than 1230*F. l i l . - _ - _- _ ,_. _ . , _ _ _ . _ , _ , ~ . . . _ . . . _ . _ _ _ _ _ _..m-,_-__.
. . . - . . - . _- . . . _ . - - . - . ~ .
i TABLE 4-1 I~ SEQUENCE OF EVENTS FOR THE SUBCRITICAL CEA WITHDRAWAL EVENT 1 Time (Sec) Event Setpoint or Value 0.0 Withdrawal of CEA's - Initiating Event -- 292.8 Core Power reaches High Logarithmic 2.7 X 10-2 i Power level reactor trip analysis setpoint, percent of design power 293.2 High Logarithmic Power Level Trip -- 3 Signal Generated 293.4 Trip Breakers Open - 293.7 Maximum Core Power, % of Design Power
- 63%
1 293.9 Maximum Core Average Heat Flux, % of Full Power Heat Flux 7.8% 294.0 Minimum DNBR 2.0 4 4 s 1 1 i 4 l t
TABLE 4-2 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE SUBCRITICAL CEA WITHDRAWAL ANALYSIS Parameters Assumed Value Initial Fission power level, MWt 2.9 x 10-8 Core inlet coolant temperature, 'F 565.5 Core mass flowrate, 10-6 lb /h 142.1 Reactor coolant system pressure, psia 1785 One pin 3-D peaking factor, with uncertainty 9.0 Steam generator pressure, psia 1178 Moderator temperature coefficient, 10-4 op /*F +0.5 Doppler coefficient multiplier .85 CEA reactivity addition rate, 10-4 4 /sec 3.23 CEA Worth on trip,10-2 5 3,79 Steam bypass control system Automatic
- 5. UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM MODES 3, 4 AND 5 WITH LESS THAN 4 REACTOR COOLANT PUMPS OPERATING 5.1 Identification of Event and Causes Refer to Section 3.
5.2 Sequence of Events and Systems Operation This event proceeds the same as the event in Section 4, the CEAW for 4 Reactor Coolant Pumps Operating. A trip is generated by the CPCs when the zero power bypass is automatically removed at 10" % power since less than four pumps are in operation. This causes the shutdown of the reactor prior to the point of adding sensible heat flux. 5.3 Analysis of Effects and Consequences Due to the prompt CPC trip at 10"4% power the consequences of this event are less adverse than for the CEAW presented in Section 3.
- 6. STARTUP OF AN INACTIVE REACTOR COOLANT PUMP 6.1 Identification of Event and Causes The Startup of an Inactive Reactor Coolant Pump (SIRCP) is presented here with respect to potential loss of minimum required shutdown margin. This event is also evaluated with respect to RCS pressure and fuel performance criteria.
Administrative procedures govern the starting of RCPs and reduce the effects of RCP starts. 6.2 Sequence of Events and Systems Operation SIRCP can either raise or lower core average coolant temperature. The average temperature can be lowered by increased heat transfer to the steam generators caused by increased core coolant flow and by colder primary system water in the steam generators being forced into the core. The core average temperature can be raised by increased heat transfer from the steam generators to the RCS as a result of increased core coolant flow and by hotter primary system water in the steam generators being forced into the core. The SIRCP event which lowers the core average temperature (the cooldown event) combined with a negative isothermal temperature coefficient (ITC) produces a positive reactivity insertion. The SIRCP event which increases core average temperature (the heatup event), combined with a positive ITC produces an increase in RCS pressure and a positive reactivity insertion. 6.3 Analysis of Effects and Consequences SIRCP can cause either a heatup or cooldown of the primary system depending on the primary to secondary A T. SIRCP was examined in Modes 3 through 6, since plant operation with less than 4 RCPs running is only permitted in these modes.
i I I 1 A. Mathematical Models i
'1he reactivity added to the core during a heatup or cooldown SIRCP event was determined using conservative isothermal temperature coefficients J (ITCs) with the maximum uncertainty applied. These ITCs were used with the maximum core temperature increase or decrease to determine the maximum reactivity inserted during SIRCP. This reactivity insertion is i compared to the minimum shutdown margin required by the technical specifications.
l B. Input Parameters and Initial Conditions The initial conditions considered for this event ranged from a positive
; to a negative temperature difference between the secondary and primary l system. Primary system temperature higher than the secondary (a positive
,' temperature difference) would result in cooling down the RCS. Secondary system temperature initially higher than the primary temperature (a negative temperature difference) would result in heating up the RCS. Cooling the RCS would increase reactivity if there is a negative ITC. Heating the RCS would increase reactivity and RCS pressure if there is a positive ITC. To conservatively calculate the reactivity added to the core . during . i l SIRCP, the most negative or positive ITCs are used with uncertainties i j applied in the most conservative direction. The initial core average i moderator temperature during SIRCP is assumed to be at the temperature i corresponding to the most positive ITC for the heatup event, or the most i t negative ITC for the cooldown event. i l The following assumptions are made: I i
- 1) Prior to SIRCP all reactor coolant pumps are off. Normally at least one RCP must be running (or one shutdown cooling train during shutdown cooling operation). The technical specifications allow
! operation without any pumps running for up to one hour. This l assumption maximizes the change in temperature during SIRCP. I . - - - - - . - . , , , . - . - . - . . _ _ - . - . - - . . - . - - - - - - . . . . - - - _ -
- 2) Following SIRCP the core average temperature either (1) drops to the temperature of the coldest steam generator, for the cooldown event, or (2) increases to the temperature of the hottest steam generator, for the heatup event. This conservatively bounds the maximum . change in core temperature that can occur during this event, by ignoring coolant mixing that would occur in the reactor coolant system.
l C. Results The results show that the maximum temperature change during SIRCP when used with the most conservative ITCs does not result in a loss of the minimum required shutdown margin. When the RCS is above the conditions requiring low temperature over
< pressure (LTOP) protection, the SIRCP event that results in a heatup of I
the RCS will not result in a peak pressure greater than 110% of design i pressure. While the RCS is in the LTOP mode, the shutdown cooling system , (SCS) relief valves will prevent violation of RCS integrity limits. (See PVNGS FSAR Section 5.2 for a general discussion of RCS integrity.) t Since shutdown margin is not lost during the event, there is no increase in heat flux and therefore no decrease in minimum DNBR. i 6.4 Conclusions I The SIRCP does not result in a loss of shutdown margin. The increase in i , pressure during this event will not result in peak pressures above the applicable limits. There is no increase in core heat flux and therefore no l decrease in minimum DNBR. 4 1 i i l
i i
- 7. INADVERTENT DEBORATION
- 7.1 Identification of Event and Causes ,
4 The Inadvertent Deboration (ID) . event is presented here with respect to time available for operator corrective action prior to the loss of minimum required
- shutdown margin. Fuel integrity is not challenged by this event.
i l The ID event may be caused by improper operator action or by a failure in the i
! boric acid makeup flow path which reduces the flow of borated water to the charging pump suction. Either cause can produce a boron concentration of the charging flow which is below the concentration of the reactor coolant.
The ID event initiated during each of the six operational modes defined in the technical specifications was evaluated. This evaluation . shows that MODE 4 (hot shutdown) results in the least time available for detection and
; termination of the event. This is because the shutdown margin . requirement which will be specified by the technical specifications is at its minimum value in the lower temperature range of MODE 4 and the boron dilution time
- constant which drives the dilution rate is also small in MODE 4. This
! combination of a minimum shutdown margin and small time constant results in the fastest dilution rate and, therefore, yields the shortest time to a l complete loss of shutdown margin.
1
- Since boron dilution is conducted under strict procedural controls which specify limits on the rate and the magnitude of any required change in boron l concentration, the probability of a sustained and erroneous dilution due to 1
operator error is very low. l ' The indications and/or alarms available to alert the operators that a boron dilution event is occurring in each of the operational modes are outlined below. I , 1 1 l, l -. - .-- -_ - - - -. . .. - - . - . _ . - - . . --
c.
- 1. The following control indications and corresponding pre-trip alarms are available for MODES 1 and 2: a high power or, for some set of conditions, a high pressurizer pressure trip in MODE 1 or a high logarithmic power level trip in MODE 2. Furthermore, a high TAVG *1***
may also occur prior to trip. i
- 2. In MODES 3 and 4 with CEAs withdrawn, the high logarithmic power level trip and pre-trip alarm, and a high neutron flux alarm will provide an indication to alert the operator of an inadvertent boron dilution.
- 3. In MODES 3, 4, and 5 with CEAs fully inserted except the worst rod stuck out and in MODE 6, a high neutron flux alarm on the startup flux channels will provide indication of any boron dilution event.
- 4. In MODE 5 with the RCS partially drained for system ' maintenance, the startup flux channel alarm will provide indication of any boron dilution event. In this plant condition, administrative controls would allow operation of only one charging pump at a maximum rate of 44 gpa. Plant operating procedure will require that the power to the other two charging pumps be removed and their breakers locked out. This drained-down case is less limiting than the MODE 4 event presented below.
The operational procedure guidelines, in addition to these indications and/or alarms, will assure detection and termination of the boron dilution event before the shutdown margin is lost. j 7.3 Sequence of Events and Systems Operation i The core is initially suberitical with shutdown margin at the minimum value , consistent with the technical specification limit. An inadvertent deboration occurs which causes unborated water to be pumped into the RCS. The resulting
- decrease in RCS boron concentration adds positive reactivity to the core.
l As sur.ing dilution continues at the maximum possible rate, 50 minutes would l elapse before the core becomes critical.
The success path is as follows: Reactivity Control: The operator is alerted to a decrease in the reactor coolant system (RCS) boron concentration either through a hign neutron flux- alarm on the startup flux channel, sampling, boronometer indications, or boric acid flow rate. The operator turns off the charging pump (s) and closes the letdown control valves in order to halt further dilution. Next, the operator increases the RCS boron concentration by implementing the emergency boration procedure for achieving cold shutdown boron concentration. 7.3 Analysis of Effects and Consequences A. Mathematical Model Assuming complete mixing of boron in the RCS, the rate of change of boron concentration during dilution is described by the following equation. dC wg = -WC Where: M = RCS mass C = RCS boron concentration W = Charging mass flow rate of unborated water dC/dt is maximized by caximizing W and minimizing M. Assuming: W = Constant, equal to the maximum possible value, i and choosing: M = Constant, equal to the minimum value occurring during the boron dilution incident, i i i i I:
the solution of Equation (1) can be written I C(t) = C(o)e * (2)
=
Where: T = M/W Boron dilution time constant
=
C(o) Initial boron concentration The time T required to dilute to criticality is given by T= T Inb- (3) C crit
=
Where: C Critical boron concentration B. Input Parameters and Initial Conditions It is assumed that the inadvertent deboration proceeds at the maximum possible rate. For this to occur, all charging pumps uust be on, the reactor makeup water tank must be aligned with the charging pump suction, a reactor makeup water pump must be on, letdown flow must be diverted from the volume control tank, and a failure in the boric acid makeup water flow path (e.g., flow control valve FV-210Y failing in the closed position) must terminate borated water flow to the charging pump suction. Evaluation of ID events initiated during each of the six plant operational modes (defined in the technical specifications) shows that MODE 4 (hot shutdown) results in the shortest available time for detection and termination of the event. Therefore, the initial conditions and analysis parameters are chosen for the hot shutdown operational mode to minimize the interval from initiation of dilution to the time at which criticality is reached. The following are the analysis assumptions for the ID event:
- 1. Complete mixing of boron within the RCS is assumed.
- 2. The technical specification lower limit on shutdown margin for hot shutdown is assumed. The shutdown margin as specified in the technical specifications can vary as a function of reactor coolant cold leg temperature. The minimum value of shutdown margin at the technical specification lower limit of temperature range of MODE 4, i.e., 710*F, is 1% A p.
- 3. The cold reactor coolant system volume, excluding pressurizer and surge line, is 12,016 ft . A conservatively low reactor coolant mass was assumed by using the cold RCS internal volume. Assuming the coolant temperature of 350*F, the technical specification upper limit for hot shutdown, the resulting mass is 667,927 lbm.
- 4. All three charging pumps are assumed to be on at their maximum rate; 44 gpm per pump, for a total of 132 gpm. .The corresponding mass flow rate, assuming cold liquid flow, is 18.36 lbm/sec.
- 5. The critical boron concentration, with all rods in except the highest reactivity worth rod stuck out, and the inverse boron worth are 752 ppm and 65 pps/% Ap, respectively, including uncertainties for the hot shutdown conditions. The initial suberitical boron concentration fcr the hot shutdown mode is found by adding the product of the inverse boron worth and the minimum shutdown margin (i.e., one percent) to the critical boron concentration. The resulting minimus. initial boron concentration in MODE 4 is 817 ppm. Thus, the change in boron concentration from 1% A P suberitical to critical is ]
65 ppm. L l I
The parameters discussed above are summarized in Table 7-1. C. Results Using conservative parameters as described above in Equation (3), the minimum possible time interval to dilute from 1.0% Ap subcritical to criticality is 50 minutes. Given the numerous indications of improper operation and the high neutron flux alarm on the startup flux channel, sufficient time is available to assure detection of a boron dilution event at least 15 minutes prior to criticality. Boron dilution will then' be terminated before loss of shutdown margin by the operator actions discussed in Section 7.2. 7.4 Conclusions The inadvertent deboration event will result in acceptable consequences. Sufficient time is available for the operator to detect and to terminate an inadvertent deboration event if it occurs. Fuel integrity is not challenged during this event. TABLE 7-1 ASSUMPTIONS FOR THE INADVERTENT DEB 0 RATION ANALYSIS Parameters Assumed Value Cold RCS Volume (excluding pressurizer surge line), ft3 12,016-RCS Mass (excluding pressurizer and surge line), iba 667,927 Volumetric Charging Rate, gpm 132 Mass Charging Rate, lbm/see 18.36 Dilution Time Constant, T, sec 36,380 Initial Boron Concentration - C(o), ppm 817 Critical Boron Concentration - Cerite ppm 752 1 l e i l i i l } l l l l r L_
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