ML20077H065
| ML20077H065 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 06/26/1991 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20077H063 | List: |
| References | |
| NUDOCS 9107050086 | |
| Download: ML20077H065 (482) | |
Text
{{#Wiki_filter:._ - . _ __. . _ - _ - _ _ _ _ _ .__ . _ _ _ _ _ _ _ _ . __ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ Attachment I ) Technical Specification changes Introduction This Attachment is divided into three parts. Attachment la contains the Technical Specification (TS) markups, Attachment Ib
-contains the Technical Justifications, and Attachment Ic contains a Nc Signilicant Hazards Analysis for each of the changes.
It should be noted that while almost all of the requested TS changes are related to the Unit 1 Cycle 8 reload in some way, not all of them may be characterized as critical to the startup of Cycle 8. Table I-l lists the TS changes requested, and briefly l describes their relationship to the reload effort. The changes i fall into four categories those required by changes in safety ' analysis methodology, those driven in some other way by the safety analyses (e. g., input assumptions), those requested to provide operational flexibility or to reduce the potential for a spurious trip, and those which correct existing errors or nonconservatisms in the TSn. This distinction between the critical and non-critical Technical specifications is offered to facilitate NRC staff prioritization of , the resources which are required for this review. While it is l desired that all of the requested changes be approved, it is . important that the review of non-critical TSs not have an adverse <
-impact on the review of those TSs which are required to restart the unit.
The attached TS are separated by unit and marked appropriately. Those pages that are not marked as unit-specific are applicable to both units. l 9107050086 910626 fDR ADOCK 03 coo 3Ap
-PDs
. ~ _ . . . . . - - . . ~
Table I-1 , Technical Specification Change Priority : C.ctocorv is loquired for Restart All changes in Chapter 2.0, safety Limits, except deletion of Neutron Flux High Negative Rate Trip and adminintrative/ editorial changes regarding the removal of references to the RTD Bypasa Manifold. All changes in Chapter 3/4.2, Power Distribution Limits. I Administrative changes to 6.9.1.9, relative to the Core Operating Limits Report. Change to Low Steam Line Pressure Setpoint (TS Table 3.3-4). Catecorv 2: Safety Analysis-Related Chances (These changes are not required for restart, but will cause the unit to be overly conservative, and possibly limited, if not approved.) Changes to ECCS pump performance requirements (TS 4.5.2 f & h). Increase in Feedwater and Main Steam Isolation Times and MSIV stroke time (TS Table 3.3-5 and TS 4.7.1.4). Catenorv 3: _ Chances to provig_e operational-flexibility or reduce the notential fol sourious trips. Removal of Neutron Flux High Negative Rate (TS Tables 2.2-1, 3.3-1, 3.3-2, and 4.3-1). Increase in Main Steam and Pressurizer Code Safety Valve Setpoint Tolerances (TS 3.4.2.1&2, TS Table 3.7-3). l Catecorv 4: Correction of errors or non-conservatisms in existinq TSs. I Increase in required loops in operation in Mode 3 (TS 3.4.1.2). Increase Cold Leg Accumulator required boron concentratior. (TS 3.5.1.1). s Change list of events to be reanalyzed (TS Table 3.1.1) i r
, .. , , , - w.,.r,y- , wv---r er -- w -
em* * ~ * ' " - ' - " * = * " ' ' ' * ' ' ' 'T ' " *-""*"
e .. l l l l f i t
)
i Attachment Ia Technical Specification Markupo
'f 1
i-l l
I!LDE,3 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY 1.IMITS((jp6_,1) 2.1.1 REACTOR C0RE.......................... ........ .. .......... 2h 2.1.2 REACTOR COOLANT SYSTEM PRE 3SURE.............................. 2-h FIGURE 2.1-1 UNITS 1 and 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS A IN 0PERATION........................................... 2 28 FIGURE 2.1-2 (BLANK)............................................... 2-3 d
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS A
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATIOl4 SETPOINTS ............... 2$4 TABLE 2.2-1* REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..... 2-h BASES I SECTION PAGE l 2.1 SAFETY LIMITS 2.1.1 B62-1 z. 2.1.2
- i. i hEACTOR
+: roc toe REACTOR COOLANT 5YSTEM P CORE.[duverMh.d.d....................
SSURE.......................... ... e p- i B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS i i 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS................ B 2-3 L A 7.b MBCHhEUT L i l
. McGUIRE - UNITS 1 and 2 III Amendment No. (Unit 1)
Amendment No. (Unit 2) l
cc . .. x o a a
- xxx Station Unit Rev File No, S h e e t .__ __ 01__ _ _ _
Subject . - - M N Nl b Dy __. Date . . . _ Prob No. __ Checked By _ _ _ __ . _ _ _ Oate_=.__ _ _ g 2.1 SAFETY LIMITS ((),o (f 2 L 2.1.1 REACTOR CORE. .................. . ... .. ................... 2-1 e 2.1.2 REACTOR COOLANT SYSTEM PRESSURE....... .. ........ ........ 24(
~
FIGURE 2.1-lh0 NITS 1 and 2 REACTOR CORE SAFETY I iMIT - FOUR LOOPS IN 0PERATION.................................. . . . . .... 2 FIGURE 2.1-2 (BLANK)............ . . . . . . ........ . . . . . . . . . . . . . . . .... 2-3
~
2.2 LIMITING SAFETY SYSTEM SETTINGS [/)dr7' 2_) 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS .......... .... 4k TABLE 2.2-lb REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..... 2-f5
, , g g e %;g-.+--> e- +-**
4, ,._,e_- 4s.,., g---qi.* ^.43'- - -
- W--*""-*W * * * *
- p. ,
, _w$ -
.-.6 49 - -'+ 6-9,,, ., , .
.m se N%w 4 6 '-4&
, p,_ ... _ ,
,,g ,.g_.., ,_,i p g. -"M"" 9 .
, , g ,4 a__ -- *
**' h * *
,.,-ag-e,-repe-$mwM----@em---M we * .
,,,_ye-.- -w- w--e. a.w h $ w-b d sr-@ -e*Oh' * ' +*
**-W *
--4~- '* -'*' "
, g-_--$--e-g---m-+% W eww-4+-
--e- - *
,m . . 4_, e. - ----
*- P
- u , ,pww.w , & -a* - -
4 + eV,e,a 9 #
,,y.
,_,,,,,,,. pac,-,.u.a . -me -m. me-==W--'- *
.,g w.p.-._.,a 4 _w.w gm.i- u--3 4 e+ W e - A
.9, ...g *. 4 4
. g. .w ws.s ,q -
..y 4 _ . .y ._ m 4 a - * - --
ee'a--+ " * *
- 9%,-,4.e-- , g. s..--4 e . -w- y
~-- ~ ~ '*' * '
- 4.-.4- = s
INDEX LIMIT!NG CONDITIONS FOR OPERAfl0N AND SURVEILLANCE RE0VIREMENT,,5 _ SECTION pAGE Control Rod Insert on '.imits.. . . .. . .. .. ., 3/4 1 21 3/4.2 POWER DISTRIBUTION LIMITS (()g,y 3/4.2.1 AXIAL FLUX OIFFERENCE.. . . . . ....... .... . 3/4/\21 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR g - F % (>( h .. . . .. 3/4 fC-6 3/4.2.3 E3 TJ at O NUCLEAR ENTHALPY RISE HOT CHANNEL ! FACTOR....... .......... .. ....... . .. . ... ... .., 3/4fC-14 3/4.2.4 QUADRANT POWER TILT RATIO. .. .. ..... . . ... ..... . . 3/4p-19 3/4.2.5 DNB PARAMETERS.... .. . ... . . . ,, . ,. ..... . 3/4A2 22 TABLE 3. 2-lo- DNB PARAMETERS.... ...... ... ..... . .. ... .... . 3/4k-23
+ @4. 3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.......... ...... .. 3/4 3-1 " FmaRz 3.2-] REAGA Coouhtrr rus) vs 2/Gsh ryc/y y,y P % %z 3 y s ~
~
' 3/4.2 POWER DISTRIBUTION LIMITS ' 't ' Q, 3/4.2.1 AXIAL FLUX OIFFERENCE...... .. .... .. ... . .. . .... 3 /4.! 2- 1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z).... .. ............. 3/412-6 9
3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.. . ................. .., .. ........... ... ... 3/412-14 3/4.2.4 QUADRANT POWER TILT RATIO.. ........ ......... ..... ... . 3/412-19 3/4.2.5 DNB PARAMETERS... ...... .... .. .. .. . ............. 3/4fE-22 TABLE 3. 2-1/> ONB PARAMETERS............. . ..... ........... . 3/t2-23 McGUIRE - UNITS 1 and 2 V Amendment No. Os (Unit 1) Amendment No.. (Unit 2)
i l INDEX LIMITING CONDITIONS FOR 0,f,ERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-1A REACTOR TRIP SYSTEM INSTRUMENTATION.((1/VII. .I. . . . . . . 3/4lh-2 TABLE 3.3-2a REACTORTRIPSYSTEMINSTRUMENTATIONRESPONSETIMES[& /4A3-9 TABLE 4.3-1q REACTOR TRIP SYSTEM INSTRUM NTATION SURVEILLANCE 3/4 A3-11 y R EQU I R EM E NT S . . (. d AJ J T. 1 . . . . . . . . . . . . . . . . . . . . . . . 3 ENGINEERED SAFETY FJA.TURES ACTUATION SYSTEM phh jw/4.3.2 3 / 4 A3 - 15 I N ST R UME NT AT I O N . L LIVIT. l}. . . . . . . . . . . . . . . . . . . . . . . . . . . .
@, TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...... ............................. 3/4 3-16 TABLE 3.3-4a
['O S /t) TA ENGINEEREDSAFETYFEATURESACTUATIONS INSTRUMENTATION TRIP SETPOINTS.(.d.M.I.T ENGINEERED- SAFETY FEATURES RESPONSE TIMES..(dNIT M 3/4A 3/443-25 J % >BLE 3.3-S a Cf TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-34
, 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS........................................ 3/4 3-40 ,
TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT
~
OPERATIONS......................................... 3/4 3-41 TABLE 4.3.3 RADIATION MONITORING INSTRUMENTATION FOR PLANT 1 OPERATIONS SURVEILLANCE REQUIREMENTS............... 3/4 3-43 i Movable Incore Detectors.................................. 3/4 3-45 Seismic Instrumentation................................... 3/4 3-46 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION........ .......... 3/4 3-47 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................ 3/4 3-48 Meteorological Instrumentation............ ............... 3/4 3-49 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............ 3/4 3-50 l TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION ' SURVEILLANCE REQUIREMENTS.......................... 3/4 3-51 Remote Shutdown Instrumentation........................... 3/4 3-52 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION..... ..... 3/4 3-53 l. 1ABLE m -)b AG/Y.TDK W WW1 WMMGUTAT1600MI'$ d 3lil 63-L
'TABis p-u RCPcT02T^lf SYSTER DJSTbHWIGT/W T
!?E.SPcw.ss ri/us (UU'T 2-) 9/4 0 Y .
[ M 43% AOcro(z 7-21f.595.TEFf .-rMsrPSmarn1s0 Yb54 \
,SLIR.Valuttq AEiptf2AEAKS (urKI2 l-McGUIRE - UNITS 1 and 2 VI nit 1)
Amendment No. $y(Unit 2) Amendment No. Q
b 3/+. 3. 2. ElJGw&:Q&b SAFerY fEnfGMS ACTJArt&) SLsrem hsrgoiswynov (uar 2J
/ 8 3 - lJ i 3' 5 '
@ TP&E 3.3 'lb Eu<;iast&b Snfer?' RersM3 ZusT RU!1EAJTartoid Tit /P SCT7brvTS (Oyrr ?k 3j'y03-u' D TABG S.3 -5~b E@IJEEfC4 SAfCf*1' FE>rTOfLS NEWscTh6s luurr7.9 sff a 3-30 i
i l l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE,qulREMENTS ___ l SECT 70N PAGE Cold Shutdown - Loops Not Filled..... ................... 3/4 4-6 3/4.4.2 SAFETY VALVES 3/442-7 Shutdown.[.d42)$.A................................. ouTPk! W T .a /4 s Ud Operating. /, d l .T .1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 44-8 _ CPCRATIU UAhT 2/4 dO 3/4.4.3 PRESSURIZER...(........................................... 3/4 4-9 3/4.4.4 RELIEF VALVES............................. .............. 3/4 4-10 3/4.4.5 STEAM GENERATORS................................. ....... 3/4 4-11 TABLE 4.4 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION..... ....... 3/4 4-16 TABLE 4.4-2 STEAM GENERAT00. TUBE INSPECTION..................... 3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....... ,..... ............. ... 3/4 4-18 Operational Leakage...................................... 3/4 4-19 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. . . . 3/4 4-21 3/4.4.7 CHEMISTRY......................................... ...... 3/4 4-22 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............. 3/4 4-23 i TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS......................... 3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY.......................................... 3/4 4-25 FIGURE 3.4-1 OOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC r ACTIVITY LIMIT VERSUS PERCENT OF PATED THERMAL ! POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
- 1 pCi/ gram DOSE EQUIVALENT I-131................ 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM............................................ 3/4 4-28 1
! McGUIRE - UNITS 1 and 2 VIII Amendment No. (Unit 1) t,.- a a n e un nonit ?)
~ - . L -
INDEX LIM _lilNG CONDITIONS FOR OPERATION AND JURVEILLANCE RE0'JIREJENTS PAGE SECTION Ice Bed Temperature Monitoring System. . . ... . .. . 3/4 6-36 Ice Condenser Doors.... .......... . ............ ... .. 3/4 6-37 Inlet Door Position Monitoring System.. ... .. .... . 3/4 6-39 Divider Barrier Personnel Access Doors and 3/4 6-40 Equipment Hatches. .. ...... Containment Air Return and Hydrogen Skimmer System.. ... 3/4 6-41 Floor Drains................................... .. . . 3/4 6-42 Refueling Canal Drains...... . ..... . .. .. ....... ... 3/4 6-43 Divider Barrier Sea 1.................... .......... . ... 3/4 6-44 TABLE 3.6-3 DIVIDER BARRIER SEAL ACCEPTABLE PHYSICAL PROPERTIES.... 3/4 6-45 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE 3 / 44 7 - 1 Safety s R.xs( dy rMrI .Q . . . . . . . . . . . . . . . . . . . . . . gyy;....... TABLE 3.7-1 q.crla ves. .iw. MAX MUM ALLOWAB_E POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION.................. ... 3/4 7-2 3/4 7-2 TABLE 3.7-2 (BLANK)........................................ .... T ABLE 3.7-3R STEAM LINE SAFETY VALVES PER LOOP. .(L/4).fC.1. . 3/467-3 ..... T w .s s. % 3 b crCAti Lab 944W &ccEs /B axe (durr %Bh5 Auxiliary Feedwater System... .......... .... .. 2 ... .. 3/4 7-4 Specific Activity................. ........ .. ... . .. 3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM............ ...... ... .. 3/4 7-7
.1 ......... . . 3/4e7-8
; J T ?_.h .9/9p -)-g 3/4.7.2 Main Steam Line Isolation Valve MAho h6AM LJ A.N butTicA) A LVf_s STEAM GENERATOR PRESSURE / TEMPERATURE LI
.. . .. 3/4 7-9 .(/2A)TT(h
....... ........ 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM......
..... ....... ... 3/4 7-11 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM..........
.. .. 3/4 7-11a FIGURE 3/4 7-1 NUCLEAR SERVICE WATER SYSTEM............
McGUIRE - UNITS 1 and 2 XI Amendment No. (Unit 1) Amendment No. (Unit 2)
l l INDEX 8ASES SECTION PAGE 3/4.0 APPLICABILITY..... ..., . .. ................ .. .. . .. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL... . ......... ............... . ..... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS............ ............ ......... ..... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES. .............................. B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS ((/gJ)f 3/4.2.1 AXIAL FLUX DIFFERENCE... ..... ................ .......... B 3/4A2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR end ECS-ILOW
-MtE-AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR. . . . . . . B 3/442-2 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/442-5 l
3/4.2.5 DNB PARAMETERS............................ ............... B 3/442-5 l
@ n -9 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION. .................... ..... ... B 3/4 3-2 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-5 3/4.4 REACTOR COOLANT SYSTEM 3/4.4,1 REACTOR COOLANT LOOPS COOLANT CIRCULATION. @.M.6 . B 3/4A4-1 Mg.a. I 3/4.4.2 kw Cawr * ^k Cwn acowsy cupT-Q SAFETY VALVES....... .... ........................... ....
o att e e B 3/4 4-2 3/4.4.3 PRESSURIZER................. ................... .... . . B 3/4 4-2 3/4.4.4 RELIEF VALVES.. ...... ... ..... . ......... . ........ . B 3/4 4-3 3/4.4.5 STEAM GENERATORS. ..... ........ ..... .. .. ........... B 3/4 4-3 l l l l McGUIRE - UNITS 1 and 2 XVI Amenoment No. 94 (Unit 1)
~ ~ ~
- .:. . :cy we Station _ Unit Rev. Filo No. _ Sheet Of .
Subject M 1 _ By Date - __ Prob No, --- Checked By_ Date .
~
_, ,,._, .,. , e e., a
-4a-4.--, Em .- -- hm-wewmm ,m.4
*w.h+h
~ - - .- ... - _ . - -
e-w--nes,_
-i-+A-.-memea -
t _ 1 .. . _ . _ . fuMT 2N_.. a-w-
- . W.L
._ OfL _dJK218vDG1L OUR -
3/4.2.1 AXIAL FLUX DIFFERENCE......,. ... . ............... B 3/482-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/432-2
.. 3/4.2.4 QUADRANT POWER TILT RATIO- . . . . . . . ........ ............. B 3/4 &-5 3/4.2.5 DNB PARAMETERS.......... ........ ................. ...... B 3t g .s
= - -h--um*N:--*:p "" -* *'
m a w e. m=s *w
, .,% g m.,,
e- - * * * = ==am -*'dF--WN'- - "4 e.eg.em'u'=r=- 4 -g euse angess= 4-e g .,.w, e.g,,,,.p.
* * ' * " " ' *~
*"*'*"* N" "**- " " ' ' '" - ' ' ' ' ' '-'
.,,gy.,,,wesae< g . .* m
,a--empir- g.eme ub. -sanmu -umme*- =., e- -
.f= mr- --
e- '- me-S-*miu.1----<$- - e- = - d --wwatmede we.suessr m :4ma .-me e-WD up-*a m-r"+- w- - --,e- 4^-6us.*>a h * * * + " - -
,,,,,. 4-e-
'~*# " ~'
.i-g-,-* engi---+ **$e=- --*M**'W'"*'N p,.,e.
' ' "*' ^
p 'd *"r**h--*""' ***##
,;,.-, .gs-s-y -m.seir.= - "w*
" ' ' ~ ~ ~ " " '***^ *' * * " ^ * * *"'~ * * ~
._e,-,.-.+ -..---um-m--ha-.ere-re~+=*--**
,e-
.em+ Na-apm@w=+-_ a vada-e----m--we-===y-ve-**"'.a
,p,,,,.
"+ *^
~ ^ ' '
'"**"5 * ' ' *' '*' '
sis -em.w.Wei -*"Wu'*'**w*W" he,-agnum,= wh r-u.s m 'e4=w n *> hw WNeuP-*E=-hN"M
'NO '*"'C ""***
summi. = wew-==ea*+rer98-*' + .
"".**e'-*4+***"#"""
4,rw ,en-4'"* hwmW- p. --6 h4%gy.gss= , . gpemen
.eM- 4" <st w'W'mmew W-"'- 8"
-.saw., e asi-am M .4.
/d ._I-LIMITING SAFETY SYSTEM SETTINGS BASES Power Ranoe, Neutron Flux (Continued)
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. fos \ T\N Power Range, Neutron Flux, High Rate / A The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. The Power e Negative Rate ripprovidhrotectionf ontrol rod d accidents, gh power, a to op acciden of a single ultiple rods Id cause loca lux peaking whic n unconserva e local DNBR to ist. The Powe ange Negative Ra could cause (l ) event this ( %m trip wil oqcurring ripping the r ctor. No creditM4 taken for eration of tAe Po h t Range tive Rate trip or those control y d drop acc Qnts for whith DNBR'b will be g ter than the dasign limit DNBR vajue. ,
- Intermediate and Source Range, Neutron Flux The Intermediatt
- and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Lew Setpoint trip of the Power Range, Neutron Flux chgnnels. The Sot:rce Range channels will initiate a Reactor trip at about 10 5 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
l l l l McGUIRE - UNITS 1 and 2 BA2-4 Amendment No,' (Unit 1) , Amendment No.e (Unit 2)
Wb , 2.0 SAFETY LIMITS AND LIMITlNG SAFETY SYSTEV. SETilNGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest shall not exceed the limits shown in operatingloopcoolanttemperature(T(EEe)t Figures 8:4=% and 2.1-2 for four and loop operation, respectively. H 2, l-lo. APPLICABILITY: MODES 1 and 2.[ ()yj f j ojyl ACTION: Whenever the point defined by the combination of the highest operating loop-average temperature and THERMAL POWER has exceeded the appropriste pressurizer pressure line, be in HOT STANDBY within 1 hour, and comply with the require-
- ments of Specification 6.7.1. i REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. (W n'th \ knd 2-) ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within I hour, and comply with the requirements of Specification 6.7.1. MODES 3, 4 and 5 WhenevertheReactorCoolantSystempressurehasexceeded2735'pbig, reduce the Reactor Coolant System prefisure to within its limit within
- 5 minutes, and comply with the requirements of Specification 6.7.1.
l
.?
McGUIRE - UNITS I and 2 2Al
. . . . . . . . _ . . .-_........-,...r ...,_,..,..,,.m., , - , . . , . , .~......~--,-.m-.....-. - ~ , , -,m..~.... ,-. -----.,--,-+.2
1 i Uld I $ h
' o. i Figure 2.11 g Reactor Core Safety Limits Four Loops in Operation (@V/T 665 FLOW PER LOOP = 96250 GPM 660 ;
655 5
- 2455 psia UNACCEPTABLE G50_g OPERATION t 645 - -
2400 psia 640 635 - 2280 psia 630 2 - c :' *
% > 625 5 .
2 .; 2100 psia g 620 - , x :
-615 ;
610 1945 psia i-
- ~ 605 -. ,
600. -. - 595 5 : g ! 3 ACCEPTABLE i -;- 590 t _ OPERATION l- : l, 585 : 580 _i i i i , , 0.00 0.20 0.40 0.60 0.80 1.00 1.20 Inac.Mn of Rated Thermal Power M0GUIRE UNITS 1 & 2
ljh f SAFETY LIMITS AND LIMIflNG SAFETY SYSTEM SETTINGS 2.2 LIMITINGlAFETYSYSTEMSETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS . i 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-ItX (/)MT APPLICABILITY: \ AsshownforeachchannelinTable3.3-1.R(UnrfJ), ACTION: With a Reactor Trip System Instrumentation or Interlock Setpoirt less conservative than the value shown in the Allowable Values colc.. of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel it restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value. McGUIRE - UNITS I and 2 2-4
Li I ABI L 2. 2- 1 CA. Cdl/ A 7 C Hl ACIOR 1 RIP $YSilli lilSTRilMEtil A110ft 1 RIP 51 IP0lHIS 5 r IRIP SEIPollif Att0WABLE VALUES f iltJC l10tiAl 1111 1 1 C
- 1. Manual Reactor Trip it . A . ii. A.
5 d 2. Power Range, Neutron flux tow Setpoint $ 25% of RATED low Setpoint 5 26% of RAiED g TPERMAL POWER lilERMAl PCWER a O liigh Setpoint 1 109% of RAILD liigh Setpoint 1 IID'E of RAllO g THERHAL POWER TilERilAt POWLR
- 3. Power Range, tientron flux, 5 5% of RAILD TilERMAL POWER with 1 5.5% of RATED It!ERMAL POWER a time constant > 2 seconds with a time constant > 2 seconds Deick liig'h Positive Rate ~
%. ~ - - . - - . - _ _ _ _ _ :
< 5% of RAILD 111ERMAl POWER with < 5.5% of RATED 11!ERMAL POWLR
- 4. Power l'ange, lieutron I lux, liigh tie 0ative Rate a tine constant 12 seconds with a time constant > 2 seconds g
a S. Intermediate Range, fleu t ron ; Zb^t of RAltu lilERMAt POWLR 1 30% of RAILD lilLRi1Al POWl R flux
- 6. Source Range, Neutron flux 5 105 counts per second $ 1.3 x 105 counts per second 1
I. Overtemperature al See flote i See Note 3 I 33. H. Overpower Al See flate 2 See Note 4 llp . i$El
- 9. Pressurizer Pressure--low ; 1945 psig > 1935 psig llp
- 10. Pressurizer Pressure--iligh 1 2385 psig i 2395 psig NP 11. Pressurizer Water level--liigh i 92% of in.,triuient ,an 5 93% of instrument span MM. > 88.8% of is.inimism mcastered I
- 12. Iow Heactor Coolant iIow g 901 of minisinum sineasured flow per loop
- flow per loop" l g: g:
23 ---
- Minimum measured flow is 96,250 gpm per loop.
-TABLE 2.2-lafContinued) AS O^>rT i
[ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS - m
- , FUNCTIONAL UNIT TRIP SETPOINT ALLGWABLE VALUES c- .
5 13. Steam Generator Water -1 12% of span from 0 to 30% of 1 11% of span from 0 to 30% of '. d Level--Low-Low RATED THERMAL' POWER, increasing RATED THERMAL POWER, increasing g linearly to 2,40% of span at to 39.0% of span at 100% of ,, , :100% of RATED THERHAL POWER RATED THERMAL POWER. . 14. Undervoltage-Reactor 1 5082 volts-each bus 3,5016 volts-each bus '; - Coolant Pumps i
- 15. Underfrequency-Reactor 1 56.4 Hz - each bus 1 55.9 Hz each bus Coolant Pumps .
- 16. Turbine Trip
- a. Low Trip System Pressure > 45 psig
, 1 42 psig
- b. Turbine Stop Valve Closure 1 1% open 1 1% open
- 17. Safety Injection Input N.A. N.A.
from ESF
- 18. Reactor Trip System Interlocks 1
yy a. Intermediate Range' Neutron Flux, P-6, > 1 x 10 10 amps 2 6 x 10 12 amps gg Enable Block Source Range Reactor Trip aa
!! b. Low Power Reactor Trips Block,'P-7 r+ r
' yg 1) P-10 Input 10% of RATED 2 9%, $ 11% of RATED THERMAL POWER THERMAL POWER q hh 2) P-13' Input- 5 10% RTP Turbine- 5 11% RTP Turbine
;; Impulse, Pressure Impulse Pressure Equivalent Equivalent-
Y({ WG A5 tUiT L TABLE 2.2-1mfContinued) ? 8 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E O c. Power Range Neutron Flux, P-8, 5 48% of RATED $ 49% of RATED " Low Reactor Coolant Loop Flow, THERMAL POWER THERMAL POWER ~ and Reactor Coolant Pump Breaker "; Position ct " d. Low Setpoint Power Range Neutron 10% of RATED > 9%, $ 11% of RATED Flux, P-10. Enable Block of THERMAL POWER THERMAL POWER Source Intermediate and Power Range Reactor Trips
- e. Turbine Impulse Chamber Pressure, P-13 Input to Low Power Reactor 5 10% RTP Turbine 5 11% 2TP Turbine .
Trips Block P-7 Impulse Pressure Impulse Pressure l Equivalent { Equivalent
- 19. Reactor Trip Breakers N.A. N.A.
- 20. Automatic Trip and Interlock logic N.A. N.A.
)
1 Allii 2.2-Iq(Continued).. N
@ HL ACIO_R I_H.IP SYSilH INS 1RUMEN_ I_ A__ll0N IRIP. Si lP_ OINIS '
, _HO.I A__1.10N '
E Q nn Null 1: OVERitHl'lRAIUHL'ai
- 68 1*t 5 al'(f,g
- 1 g ) (1 , g35) $ AI o IKg.-Ky (y , 4g )[lf g, g)-l' ]
- K3(P-P' ) - f (al)) 3 n
Winere: AI = Measured al by HID,*t.. h,:J L. & r 'c , I+t i$ L " H ## '" b " 4 3., g
= lead-lag compesisator on measiired al, I i, t2 = Iime constaints utililed in time lead-lag centroller for ;
AI, si > 8 sec., r: < 3 sec., ~ ru
=
1au compesisator on. measured al, j i3 = Iime constants utilized in the lag compensator for ai, r3 $ 2 sec. f AI, = Indicated Al at RAILD liiERMAL POWER,
- 1. % l.1958 K,
![!{ ' *i K = ; 2 0 . O '51'{ 3 ' 'b 1 . i,s 5
= .
, g 3,y 1he. function generated by the lead-lag controller for I,,g dynamic compensatsosi, em i,
i ., = lime coinstarits natilized in tiie lead-lag coritroller f or I t, > 28 sec, 1 3 $'4 sec., #"U, , G2 i'
- j. j. I =
Ave ragt- temperature, }, U.C I =
- f
_ g , ,7 lag compesisator on measured ( , i r
t f0 \. IABlE 2.2-IslContinued)' c> g REACIOR 1 RIP SYSitM INSTRilMENTATI0il IRIP StIPOINTS rn . .. M ATION (Continued) C E N0lt 1: (Continued) is = lime cosistasit utilized in the sneasured I lag compensator, is 2 sec g' I' ='
$ 588.2f Reference T avg at RATED IllERMAL POWER,
~
K = 3
, O.oo t yog !
P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure), i S c- taplacel transform operator, sec 8, h and f (AI) is a function'of the indicated difference between top and Lottes detectors i of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for qt ~4b between and 47.0% (al) = 0, where q and q are b percent RATED {' , p IllERMAL POWER in the top and bottom halves of the core respectively, and q t a q i b N is total TilERMAL POWER in percent of RATED TilERHAL. POWER; l(lh [g imbetonee for each percent"that. the magnitude of (q"J#* ^*@*-397 6:I o e (ii) - qptaceeds-J37.,theATTripSetpoint ' _._, shall be automatically reduced by of IL mie.;.G;20 !ilarmt ruacR; and
,2r ji imbt., .l537, ATo n"'o* * *
- for each percent."that the magnitude of qt ~ 9b ex:.cedd+7.0%g the AT Trip Setpoint (iii) I
""' AL -
shall he automatically reduced by % of 4 L . u i ui. et naecu autnrm euwca. GG 1 5117. 4% R R o,+ Y
[ - i L II ad!!I[ L 2:1 gun! pines!} : g .let. AC .l .i.l.H. '.lit 11'.
. _ !.Y. S_ .il. !..! ~..I N.51. RitMi .H.. I_A I .I.ON. .I.H. .i.l' :,1.i t'0. .I .N. .I- $ 't i
5, N01
--- A.l .l.0N. ' .( Cont .i n.nesi.)
m-
..- 18011 2: UVLHl'UWlH al E '
3S- 1 1 ;
*~i
~
-K cn a 1. (I1*1
' 5285) ' (I
- 1 35) < al o-{K4 5 (1 + 115) (1 + IsS) 1 -K6, [1(1 + t SJ- I"] - t2(al)) ;
.Where. Al - As 'elet ineil in flute 1, i
. a i 0 : 1*i$' i m' j7-j- = As defineal.in Note 1 ' , I i.'42 -= As.detined in Note 1
- g. ,
g ,-
- As - del luct! in thet e 1, n al " = 'As detined'in Note 1, g .., j 6
K
- % ' t. o so.9 -
4 f
=
K,3 0. 02/l for increasing average temperature and 0 for decreasing avteage temperature,
- j
p -g = c1heompen:,afunction t lon , generated by the rate-lag controller for I""9 dynasitic
.. +
i ,. i
! l :j l a, '-
lime i.unatant utilized in the rate-l.sg t.ontroller for I dv9, i , , S set . f i i
] I ; ; j ,- 3 A:. def ined in if sie 1,
- e ~
[a~<a Q tu- ' As detineal in Note 1, , y K g - p/i . tur I . > I" and K 6 = 0 f or i I" - Ti ; O.o0123g
.o . i
, s e s' s 1 i e S.
, a e,., ., ---. , . , . . .,
' I 7I , , .=~ 2 . t! tf; i!lP i .b , i-f.'*;[
!(sl, i > 7It 4 F ieh lc'< , ,
AJ L Z s)^ n4. g h n a t m i h n a t e t . e u e r r r o o m - m y ; y b L b t t t. i n ; i n o p o o p p t t e e S 'd. 5 R p p p S E i . i I N W O T r T r . I P O d d d P L e c e l A t t t i M u u S R p p E m 4
- m P i l
o ; o I R T c . c
) I D s ; s d E t i t e N )' T i i u 0 d A n 1 e R d d d i
t 1 A is n e c e n t a e ; e o I l i c ; c l t d x m x C i n g n e c e ( lM o v a i C a t' i t - 1 R. ( , I . o o n w 2
- 1.
S- l t 1 e 1 n N- 0 e c e l l l 2 I- l t n t l ; l t l . l A H o e r l o h a _ a' l f ; h i- i. 1 e s 2 s B- l. 0 n f n A- S- 1 1 i e i .n t~ d t I- Y. R .u n , n S- d d i , i e f e .i o u o P- n " n' .s. p g p ' I. i 2 i a t t R-I. t e 8. t e s. S e dL h e d 8 d' u r . R'- 5 . a pe g p 0- s s' i y ; i 1 A <- A' u ro , r' t I P 1 f A- l r f-k-
= =- =E:. - ml ua o
u me uw n mm u imo s, ir P s xe r, x ah a. al m mT u mam M,,..' e s sd s sr
" t ' ' e I I S l t l l h ea eT nR n t, n il t i
- r. af ae 5
e h o. c ht
<a 6 % R e6 e 1 h .
I 3 h hf o Hl .
- a .
3 a 1 e c e t o d t i h h f l t I
. goc
- gm , cz m " moa.*
. .l1l Lt lt .i l s:s: .- A Y'3r E#ICIe:r W(. 'i : t%
C vC: I>:: _ m.11lkli lP l h$ _ :#!t?tlr mC. (- E
,]! * , i ..! .'
5 4 4 E *
. . .- . ... .. . . - - . - _ . . - . - - - - - . - . . - - .- .. .~ . . .. .-
-Attachment 1
. and f,(AI) is a function of the indicated difference between top and .
- bottom detectors of the power-range nuclear ion chambers; with gains to be' selected based-on measured instrument response during plant star. tup tests such that:
(i) for qt .q, between -35% and +35% AI; f,(AI) = 0, where qt and + q,are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + q, is total THERMAL POWER in percent of RATED THEPMAL POWER; (ii) . for each percent imbalance that the magnitude of qt - go is more negative than - 35% AI, the AT Trip Setpoint shall be automatically
- reduced-by-7.0% of AT,; and (iii)- for each percent imbalance that the magnitude of qe - q, is
- more positive than +35% AI, the AT Trip Setpoint shall be automatically reduced by 7.0% of ATo.
t h 1 l l c., - - --
., - - . . . + - .n.,
TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENI 0; AN INOPERA3LE FULL-LENGTH ROD Roc C:uster Control Assemoly Insertion Characteristics V6sootratson Roc Cluster Control Assemoly Ri;;iign.;ent Loss of Reactor Coolant from small Ruotured Pipes or frem Cracks in large Pipes whicn Actuates the Emargency Core Cooling System _5 '-f e ce* "'" ster Co. L v; A>>emaiy eitaarawai ou Fv;; ,'c.er e sjcr 9tcct;r Cocisni. Cy>i.cm Pipe Rwtttrc 3 ( Les s cf Cas44nt, at:: centr Major Seconcary System Pipe Ruoture Ruoture of a Control Rod Drive Mecnanism Housing (Rod Cluster Control Assemoly Ejection) I McGUIRE - UNIT 51 ano 2 3/4 1-16
(dNif / 3/4.2 DOWER DISTRIBUTION LIMITS 3/4.2.1 HIAL FLU _X OIFFERENCE ( AF01 LIMIT 1NG c:NDITION FOR OPERATION 3.2.1 _ a cce p+. ble Thetim.h indicatea as o.c;4.J AXIAL
;w % cm FLUX ow.s,OIFFERENCE Limas fWt (cota)(AFO) shall be ma f a.
REPORT (COLR) for RAOC operation, orthe allowea coerational space c
" a
% b. {
the target band loadspecified operation.in the COLR about the target flux dif ference { during Dase $ APPLICABILITY: MODE 1 above 50% of RATED THERMAL ( W:4- POWER
- 1) *.
ACTION: OJ.I. a. For RA00 operation with the indicated AFD outside of the limits soeci ed in the COLR, C 1. Q) Either restore15 limits within theminutes, inoicatedorAFD to within the COLR 2 2. Reduce THERMAL POWER to less than E0% of PATED THERMAL PO within 30 minutes and reouce the Power Range Neutron Flux - High Trip setpoints to less than or ecual to 55% of RATED THERMAL POWER within the next 4 hours. For base load operation above APLND" with the indicated AXIAL FLUX O!FFERENCE flux difference: outside of the aoplicable target band about the target ( bds 1-Either restore the inwicated AFD to within the COLR specifiec ' s target bana limits within 15 minutes, or k 2. ( Reduce THERMAL POWER to less than APLND ano discontinue Base Lead operation within 30 minutes. of RATED THERMAL POWER
- b. c'.
THERMAL POWER shall not be increased aoove 50% of RATED THERMAL POWER unless the incicated AFD is within tne limits specifiec in the COLR. 6d
'See Soecial Test Exception 3.10.2.
NO
**APL f s tne minimum allowaele (nuclear design) power level for base loao operation anc is specified in the CORE OPERATING LIMITS REPORT per Soecification 6. 9.1.9. &
A*Jb L McGUIRE UNIT 31 afro-2 3/M 2-1 *n nt No. MC (llnit I) E.. . n_ . n. .Nv.67 swuii O
Nfs T} 00VER DISTRIEUTICN L:MIT5 EURVEILLANCE REOUIREPENTS
- 4. 2.1.1 30WER OPERATION acove 50% of RATED THERMAL 70kER by:The
- a. .
Monitoring the indicated AFD for eacn OPERABLE excore enannel: 1.
' At ano least once per 7 days wnen tne AFD Monitor Alarm is ;PE?>BLE, 2.
At least once per nour for the first 24 hours af ter restoring the AFD Monitoring Alarm to OPERABLE status, b. Monitoring and logging the inoicated AFD for eacn OPERABLE excere enannel at least once per hour for tne first 24 hours and at least once oer 30 minutes thereafter, when tne AFD Monitor Alarm is inoceraole. The loggea values of tne indicateo AFD shall be assc:ee to exist during the interval prececing eacn logging. 4.2.1.2 The indicatec AFD shall be considerea outside of its liuits when at least two OPERABLE excere enannels are indicating the AFD to ce cutsica tne limits. 2.1. 3 When in Base Load operation, the target axial flux cifference of eacn CPERABLE excere enannel shall be determinea by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicante.
- 4. 2.1. 4 When in Gase Load operation, the target flux difference shall be uccatec at least once car 31 Effective Full' Power Days oy either catermining i the target flux cifference in conjunction with the surveillance recuirements #
' of Soecification 3/4.2.2 or by linear interpolation oatween the most recently i measurea value anc the calculatea value at tne ena of cycle life. The orovistons 3 t
i of Scecification 4.0.4 are not applicaole. w: 5 [Oclt w J l 2 gA1
- l. u GUIRE - UNIT! 1 M c ^ 3/4 2-la 'r:= C2 b o - ?
wn,. .nr n e r ('y ". . . _
. . . . ~ . , - _ . - -- - - . - . .... .-. . . . - .. .- - --
UMT1-POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg . LIMITING CONDITION FOR OPERATION (4 jpQ%lhk 3.2.2 p shall be limited by the following relationship: Fp 5 F QRTP K(Z) for P > 0.5 X.,
$ RTP N1 I 0. 5 Q K(Z) for P 1 0.5
- Where FRTP = the F limit at RATED THERMAL POWER (RTP) specified 9 9 in the CORE OPERATING LIMITS REPORT (COLR),
_ THERMAL POWER , and
~ RATED THERMAL POWER [//1[
K(Z) = the normalized N F ff7 W ivea cor: hew 3 specified in the COLR, Fag -rlJe APPRef[/AQ APPLICABILITY: MODE 1. (UAl6 3 O/6(. 7Q N f(q l ) yl p g p}$()] Wity exceeding its limit: fb
- a. THERMAL POWER at least 1% for each 1% F (Z) exceeds the i t nutes-andsimilarlyreducethePowkrRangeNeutron Flux-High Trip Setpoints Dwi Mi d th Qnext 4 hours; POWER OPERATION may proceed for up to a total of 72 houTsds uent POWER OPERATION may proceed provided the Overpower 0 Setpoints (value of K 4) have been reduced at least-1% (in AT span ach
- 1% F (Z) exceeds the limit; and
- b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the-reduced limit required b /
ACTION a,, above;' THERMAL POWER may then be increased provided - h$ is demonstrated through incore mapping to be within its limit. W [Q x)lJ-J : ff}$ 06MUfbh NSAK ZUh h0YCSkMt5L Q fojk ( % f W lI $ C S 9EC 1416b jid 4. L 2 Q McGUIRE - UNITS 1 and 2 3/4/t-6 Amendment No,1Y5(Unit 1) AmendmentNo/SUnit2)
for specification 3/4.'2.2. Attachment'1:
- a. Reduce THERMAL POWER at least-1% for each 1% F7(X,Y,Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip retpoints within the next 4 hours, and b.
Control-the AFD to within new AFD limits which are determined by reducing the allowable power at each point along the AFD limit
- lines of Specification 3.2.1 at -least 1%- for each- 1% F7(X, Y, Z) exceeds the limit within 15 minutes and reset the AFD alarm setpoints to the modified limits within 8 hours, and c.. POWER OPERATION may proceed for u'p to a total of 72 hours subsequent POWER OPERATION'may proceed provided the Overpower AT Trip Setpoints (value of .K.) have been reduced at least 1% tin AT span) f or ' each 1% F7(X, Y, Z) exceeds the limit, and
dne POVER DISTRIBUT:CN LIMITS SURVEILLANCE REOUIREMENTS 4;2.2.1 The orovisions of Scecifkation 4,0.4)are not apolicaole. Af' C s,1, t ) \' 4.2.2.2- E r M 00 . e.e u en,( (F (z wLag, - ( h lW ) is within_its_ limit by: n snall be evaluated to cetermineuf (g a. Using the movable incore cetectors to ootain a power distribution mao at any THERMAL POWER greater than 5% of RATED THERMAL POWER, 6 Increastng the measurea 9F (z) component of the power distribution mao by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, Verify the requirements of Specification 3.2.2 are satisfiea. f
- . Satisfying the following relationsnio:
,r ~ . Dste+c 's RTP
% cM F g -
' Q (I) c- P ,xg2) for P > 0.5 -
w(z) RTP M F g (z) "O - w(I) xv0.5Kf') for P < 0. 5 wnere F (I) is the maasurea F q (z) increased by the allowinces for manuf acturing tolerances and measurement uncertainty, 3, gn, Fg limit, F)RTP 7 K(z) is tne normalizea F g (z) as a function of core neignt, P is the relative THERMAL POWER, anc W(I) is the cycle.decenoent 2 function-tnat accounts for power cistribution transients encountarea during normal operation. F g , K(I), and W(Z) are specifiec in the 5 CORE OPERATING LIMITS REPORT per Scecification 6.9.1.9. 7 l c. Measuring gF "(z) accercing to IrIIolic, wing senecule: 1.
- ~ Upon achieving eauilibrium concitions _ af ter exceecinq cy i'fk" 2 10% or more of RATED THERMAL POWER, tnr THERMAL POWER at
*nich FQ(I) was last cetermineo," or o w.d ..,uj V , 2. At least once per 31 Effective Full Power Days occurs first. nienever f'
"During power escalation at the beginning of eacn cycle, power level may be increasou until a power level for extencea comration has been acnievea s ano a eewer cistritution mao cotaineo.
McGUIRE vNITS 1 =:
' I Ms 7 3/42-7 ?
. - nu nodeiun. L U-
']
.e... . .~ yn g n. .s .s. .3.. .. n. e. m......
n..,......
.. 3.3.e.... . . . . . ,, .r ;r
* , A" . v. , . ) 3 *. * *...e e a . .' .' *. a" *. .~#.* .
.' a. 3 5 *. .' ". C ey *er .I .' ? .# .# .S" *. .* ". e .T .' . . ^^We.*
.. A 'w . . v ., 'ayS,
.. .. . c, o n .- a 5 . . . . . . g e q'r. .' .' '... , -.- " .. . , a... c .' *. . .- a. 5 a .f . a. .* a,. x c e e c .. . , .. ;
. '.'.t.
.. . . .o .,. e ,. ., .m-- . u. . r. P.v.A v. c- ~a~c.dn , . . e +. u.r r. val
. ., t
. . . . . m .R cont a. .. .....
T". ( X , Y , 2 ) was last ceter:"inec ', :r n, ,. s. . n .. . . ...e
.w
. . . e ,. n.w" Fs' ".. '". 'i.^ W E R .. " ". e. s' ". . ^. .a. '. *. .ar.e*. .-. - .=.
.a .e. . . . . 3.i .,00 ...,.e aet.,. ...
.XCw.e
. . ,. . eC........ . . . . . , ..,.. ...
" e & S u r e.".e n t 3 .
s ., J
. . . . . 3., .... .g
.. 3 . ...a3. . 2. . .,,.. 3. 4
.. . . . . .s ..
. t ., .o w ., p. ,y
- a. m, .. .s. ...
...... e., ..
. .r*; ( X , v. . *. ) ." O C S" S e * .*. *. . . ' . ' . . . .*** .'.~.~..'"".'O '.'.".*e"*..'..'.*.*..'*..*.
t S.
- s. )
~ . . . . . . '
...y *, *. W. .e . 2. S *. a .l a *w .4
^*
... 3 *. . *** . . . "* C_.'.*..*..**..* ^# .. P. 3 C .~". ; C .l e ,. ~..U....'..'.
- . v. c .~. "'ay O e 1.". C r e a S e c '.; n t i . a pCWer .3 Vel IOr eXtenOOC O p e r a t i!"-
. ., S .n. o .n. 2.... . m.. y ,. g 3.p 3 -y.w p. .. .........n. a..'. . ~ ., y . ..3 .e
Ynk f 30VER O'57:15UTI*H U MITS IURVEILLANCE ?!OUIREWENTS (Snt'-';eo)
- , t .~. '
~^
[e. .itn tessurements ino1cating maximus F;
- )
Over : M :; > as increason since tne crevious , ins felicwing actions snail be sat inaston of F)9(I) ettner Of
*)
. : Q*(:) snall be increasea by 5 cver that soecifica in Scacifi-
{' =atten 4. 2.2.2c. sr
- d
) g (:) snail be measurea at least once car 7 Effective Full I Power Days until two ..:cessive mass indicate snat H
maximus ( F'O (:) is not increasing. over : ( s(I)
- f. ,ttn :ne relationsnics scactfino in 5:acification 4.2.2.2c. aoove ot : sing satisfiaa:
*)
lalculate tne carcent Fq (:) excenas its limit ty the following expression: { maxtmum l 0 'g) , ggg) } b p Me 100 for a > 0.5 (over: Fg g - j
; x K(:) s g
m (Imaxtmus p M,, Qty, a 0 ") ,g(*) for P < a.5 escatwwe 3 - (qay,7: : W. 3
)1[x100.
-g
_ > a K(I).) / ' d L Jy / 1 l
- 2) One of the following actions snall be taaen: I a) dithin 15 minutas, control taa AFD to within new AFD limits
.nien are caterminea by recuctng the AFD limits of Soec1fication 3.2.1 by 1% AFO for eacn cercent F g(:) excenas .
l its limits as caterminea in $cecificat1on 4. 2.2.2f.11. ditnin 8 nours reset the MO alarm setecints to tnese utfica lim 1ts. :r i
- ) Ocamly with tne requirements of Scecification 3.2.2 for 3; ( )
l ( exceecina its limit tv the carcent calculateo aoove,_:r
, :)- sertfy tnat tne recuirements ef Scactfication 4.2.2.3 for -
-%,% ~
ase loac oceration are sat 1sfied ano enter case loao coerett en. t t l i 1 I
'cGUIRE .*t!T1 1 *p *e-++ 3/4A2-8 ' ----- u . G L i..i -.
*------m anunts 2)
. . _ . _ . _ _ _ . _ _ .. _ _,-_____m _. _ . _ _ . .____..m...__.__.__..______-,__..._.._
f:r Spec;fication 4.2.2.2 Attacncent 3:
- Perf:rming the following calculations:
- 1. For eacn_ core 1: cation, calculate the 4 margin :: the max; mum allowaole cesign as foll ws:
Fj (X, Y , ;)
% Operational Margin = ( 1- ) x 100%
( Th (X, Y, 2) l Fj(X,Y,0) ( RPS Margin = ( 1- ) x 1")0%
. s. 2 ( X , v.. , ) 1 n where (Tj tX. Y,0) ]" and rt ("., t. :'.) J '" are the Operational anc RPS '
design peaking-limits cefinea in the COLR.
- 2. 71nc the minimu:r Operational Margin' of all locations examined in 4.2.2.2.0.1 acove. If any margin is less than tero, then either of the :following actions shall be taken:
(a) Within-15 minutes: (1) Ocntrol the ATD to within-new AFB '.imits that are
'determineo by:
~*-*R i3 M ::
(AFDLimi1-[eouces e g a t ., ..e
.(AFD Limit 1 .!."e g" a ,. .,v.ARG ,'.,e :!! . . 3 (AFD Limit) ec [* s"t#N %RG:iY.k"
= fAFD Limit! pas.... ve -
t..e c v Where MARGIN wP
%nt ;s Ene minimum ma rgin '::m 4.2. 2. 2. 0. . .
and . (-)- ;;itnin 3-hours. reset the ATO 5.a = setpoints e mocified limits of '4.2.2.2.c.2.2, :: (c) _ C:mply' with the ACT 2:1 requirements -:t Specificati:n 3.2.2, treating the margin violation in 4.2.2.2.c. '. acove - as tne amount by wnica Fi^ is exceecing ;.3 .1= t.
- 2erinea anc specified in the- COLR per Specificati:n 6. 9.1. 9.
_ . _ ~ . . . - , _ .. , - -
.e n,. e. e n. c ..a. . ., .. .. ,
. n. *.c..~,.-
. ~
At t ach: .e n 3 (:Ont)-
- 3. Tind the min:. mum FJS Margin of all 10 cations exarnned in
- 4. 2. 2 . 0. 0.1 above . f any margin is less tnan :ero, then :..e foll wing act:On snall be taken: ,
Within 72 hours, recuce the K: value for CT.iT by:
.. adiusted -K,(4) min,
- 5. K5 w, ,,P E ( 3 ) x Margin n; -
gp3, absolute value n. wnere MARGIN 'f'7' c.P S
- ,s the minimum margin frcm 4.2.2.2.c.1.
l-l Oerans: an specifie: 1.. ne C2LR per Spectficati:n 6.9.1.9.
" K. value 'f rom Table 2. 2-1.
i l
. . - -._ - .-.=. . . - . - - . . _ - - - - --_ . - - -. . ._
for Specification 4.2'.2.2 Attachment 3 (cont) :
- d. Extrapolating"' at least two measurements to 31 Effective Full Power Days beyond the most recent measurement and if:
(FE (X, Y, Z)' ] (extrapolated) 2 ( Fj ( X, Y, Z ) 10' (e xt rapolat ed) , and f F5(X, Y, Z) 1 (extrapolated) > I F5 ( X, Y, Z ) 1 ( Fj (X, Y, Z) ]" (extrapolatec) [ Fj (X, Y, 2) ]"
'or
( F0 (X, Y, Z ) ] (extrapolated)- 2 (Fj (X, Y, Z) ]"8 (ext rapolate d) , and IF5 (X, Y,2) 1 (ext racolat ed) > IF5(X, Y, Z) 1 ( Fj (X, Y, Z) ] * (extrapolated) (Fj(X, Y, Z) j
- either of the following s -lons shall be taken:
- 1. FE(X,Y,Z) shall be increased by 2 percent over that specified in 4.2.2.2.a, and the-calculations of 4.2.2.2.c repeated, or
- 2. A movable incore detector power distribution map shall be obtained, and-the calculations of 4.2.2.2.c.1 shall be performed no later than the time at which the margin la 4.2.2.2.c.l_ is extrapolated to be equal to ::ero,
"' Extrapolation of F; for the initial flux map taken af ter reaching equilibrium conditions is not required since the initial flux map establishes the-baseline measurement for future trending Also, extrapolation of 'Fy limits are not valid for core locations that were previously rodded, or for core locations that were previously within-22% of the core height'about the demand position of the rod tip.
4 1
-4. . . . - - , _ . , - .,,.
. - - - _ - _ - - . - . . . - . - . =. .- - - - . - -- U t&t- l DOWER DISTRIBUTICN LIMITS SURVEILLANCE ltEOUIREMENTS (Contt r ueo) [g. he limits specifiea 1,
' w-M Soeci fications 4. 2. 2.2:. t. 2. 2. 2e. , ana 4. 2. 2. 2' '\
u '~ Y , aoove are r.ot acclicaole in the folio.ing core clane regions: 1. Lower core region f rom 0 to 15% inclusive. U Q 2. Upper core region from 85 to 100%, inclusive. ( 4.2.2.i ::ase ioao operation is cernittea at cowers aoove APl# if the j following concitions are satisfied: a. Prior to entering case loaa operation, maintain THERMAL POWER aoove I ( APL j ano less than or eoual to that allowea by Scecification 4.2.2.2 for at least th: crevious 24 hours. Maintain case load operation surveillance (AFD within the target band about the target flux ' dif ference of Specification 3.2.1) during tnis time period. Base "' loac operation is then permitted providing THERMAL POWER is
'- maintainea between APLHO and APL O ND
'gg or between APL and 100% (whicnever is most limiting) ano FQ surveillance is maintainea cursuant ,l T, to Soecification 4. 2.2. 4 3
' -r - APL ' is cefinea as: f
\
RTP l 3pt3L , minimum g (F0 x K(Z) lC
#V'" i x 100% b
}
F' (Z) x W(Z)
.nere:
F0 (z) is the measured F9(z) increasea by the allowances for manufacturing tolerances and measurement uncertainty. RTP F is the f g ? limit. 9 ((z) is the normalize.c Fn (z) as a function of core heignt. W(z)gg is tne cycle decencent function :nat accounts for limiteo '
.f cower distr 1oution transients encountered curing case load coerat10n.
F ,
'(z), anc W(z)9L are specified in the CORE OPERATING LIMITS I REPORT cer specification 6.9.1.9. '
/ o.
During base loao operation, if the THERMAL POWER is cecreaseo below [ APL 0 re-enteringthan the case concitions of A.2.2.3.a snall be satisfiec before loac operation.
. a. 2. 2. 4 F,(Z) is within During case Icao aparation gF (Z) snall os evaluatea to cetermine if s
its limit'Oy: a.
. Using the movaole incere cetectors to cotain a cower cistribution ,
4 mao at any TuERMAL POWER above APLND o. Increasing tne measurec F g (Z) comoonent of tne cower distribution mac Oy 3% to account for manuf acturing tolerances anc further increasing tne value ey 5% to account f:r measurement uncertainties. T Verify tne recuirements of Specification 3.2.2 are satisfied. ;
/
?
" AP L.10 l is tne minimum allowaole (nuclear cesign) ower level for case loaa operation in Specification 3.2.1. ,
S McGUIRE - UNITS 1,Aub me-s 3/442-9 ' r 1_.. L.iCTJm 6 -,
-i s
e_ec,.,
. . .s...., . . . .
n..,.. .... . . . . . . . . . . , . ,.
- s. ~'he .imit s :.n S peci f t :sti:ns 4. 2. 2. 2. 0 and 4.2.2.2.d are .:.
- .". e f ol?.cwang c:re plane regions as reasurec :n appli:scle, 1..
w ...e.,
. .... ...e m
. x e. . .. . e....e ... .e.. . e: ..
ee e ., .. . . . o . .. ., .e . . e 1,.,, . . .
- 1. ' ower ::re- region f:cm 0 to 15L inclusive.
f-
- 2. Upper core region from 35 to 100).-inclusive.
U. - f.
- 0WE: ::ITRi30TICN L:MITI IJRVEI'.1NCE :EOUIREuN! '":nt' tee t
- . !attsfying t..e f:ll:=ing reiat1:nsn10:
K
.ATP C
- U(-)<-
4 9m <:r a> apt NO d 3t 9 RTP S
! .nere: FQ (Z) is the measurea F.(!). F is tne F Q11mit. C w Q
((Z) is the normalizea F)(!) as a function of c:re neignt. : is tne f elative THER.uAt ;0VER. i
\
4(Z)3t is the cycle cecancent funct):n tnat
\
j accounts for limitea ecwor cistribution transients encountereo curing case loao courat1:n.
~
, X(I), ana.W(I)gg are scac1 fica in the !
l0RE CPERATING L:MITS REPORT :er 5:acification ti.9.1.3. t 9
-- :. deasuring wF,(;) in c:njunett:n with target flux 01fference ceter -
21natten acc:retng : tne follcwing seneaulo: ;
"r -
t 1. Drior t: entering case loaa coeration af ter satisfying Section s v! 4.2.2.3 unless a full core flux man nas caen taren in tn
- revious a,1 :...,0
.. r with tne relative thermal ::ver naving :een ;
I i maintainea aoove APL'10 for tne 24 hours price to maccing, ana .
- 2. At least :nce car 31 effect1ve full power cays. l
- e. dita measurements inoicating i; F'(2)
}aximum
.ver -
( g ., M ' mas increassa since tne crevious cetermination F.(!) eitner of tne
'Ollcwing actt:ns snall to tamen: *
'.. r 9g(!) snall be increassa ey 2 :ercent over .at scacifito in 4.2.2.4.c. Or
- . r Q(Z) snall be measurea at ' east Once :er ' ???O until successive maos ' :1cate trat 9
i F3 (!) maximum ( %w,)-- : is net increasing. Over :
.ith tne relat1:nsn1c s:ecifiaa in 4.2.2.4.c acove not :eing sat 1sfiea, ettner :f t .e f ollowing actt:ns snali te tagen: ;
. : lace tne c:re in an ecu11br1cm concition wnere tne limit in
' 9
- 4. 2.2.2.c 1s satisfiaa, ana remeasure F)(Z), :r
% l
#A/6 2.
dcGUIRE UNITS 1 ane-+ I/442-9a
~ _____.',
- * " " " 4 i ' '-
- u. asw v g1s
1 jn(" 20WEP OISTRIBUTICN LIMITS SURVEILLANCE REOUIREMENTS (Continueo s f
,/ 2. Comoly with the recuirements of Specification 3.2.2 for Fg(Z) -s'
/ exceecing its limit :y the percent calculatec with the following .
expression: f }
-t
[(max, over : of ( 'f(Z) x W(Z)E ] ) -1 ] x 100 for Pl >g,, API.ND 7.v P l / g( , f x K(Z)
! g. The limi ts speci fied in 4. 2.2.4. c , 4. 2. 2. 4. e, and 4. 2. 2. 4. f above :
are not applicania in the following core plan regions: l
- 1. Lower core region 0 to 15 percent, inclusive.
- 2. Upper core region 85 to 100 percent, inclusive. -d 4.2.2.5 When F (Z) is measured for reasons other than meeting the recuirteentD of specification 4.2.2.2 an overall easured F (z) shall be obtained fec= a power 9
j cistribution man ano increased by 3% to account for manuf acturing tolerances (ano further increased by 5% to account for measurement uncertainty.
\
[
'tf A4.s Wh MadM r w
f
*cGUIRE UNITS 1 ano 2 3/4A2-ob A - ~rit '010M 5i; U L ou..ns av,oi(UniL 2)
Attachment 5: 4.2.2.3 When a full core power distribution map is taken for reasons other than meeting the-requirements of Specification 4.2.2.2, an.
,- overall F5(X, Y, Z) : shall be determined, then increased by 3% to account for manufacturing tolerances, further increased by 5% to accountifor measurement uncertainty, and further increased by the radial-local peaking factor to obtain a maximum local peak. This values shall be compared to the limit in Specification-3.2.2.
4 5 6 i e ~ l l-
UutT 1 POWER DISTRIBUTICH LIMITS 3/4 2.3 !! ~' C'_' :."T"
";C NUCLEAR ENTHALPY RISE WOT CHANNEL IACTOR - F39 [x y)
LIMITING 0 NOITICN FOR OPERATICH 3.2.3 The c moination of inoicatea Reactor Coolant System (RCS) total flew rate ana R snall be maintainea within the region of allowaole operation ) i soecifieo in the CORE OPERATING LIMITS REPORT (COLR) for four loop operation: L, Where: N F
- a. R=
r
.RTP
' Mi (1. 0 MF g (1.0 - 0)] ' {
Itf u c8 k 8 = THERMAL POWER ATTN:N b. , RATED THERMAL POWER i.
- c. = deasurea values of F cotainea by using the movaole intore catectors to cotain a power cistribution map.
f l The measurea N
, values of Fg shall be usec to calculate R since the figure I specified in the COLR includes penalties for undetected
( feeawater venturi fouling of 0.1.% and for measurement uncertainties of 1.7% for flow anc 4% for incore measurement j , N cf Fj, I
; d.
~
' :The F" limit at RATED THERMAL POWER (RTP) specifiea in the f COLR, ana '
)
m
- e. MF', = The power factor multiplier specified in the COLR.
l ,.- -- - - APPLICA8ILI*Y: %0E 1. (Ut'i 4 I) ACTION: s 4 With the cc:noination of RCS total flow rate anc R outsice the region of acceotaole coeration specifiea in :ne CCLR: i,
- a. within 2 hours either:
.'. Restore tne comoination of RC5 total flow rate ano R l
y to within the aoove limits, or l PGLLCE i i Ni 74 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER g pgg , anc reouce the Power Range Neutron Flux - High Trio Setooint , I q to less than or eoual to 55% of RATED THERMAL POWER within / the next 4 hours. / l
~
icGUIRE .h!T~ '. :ac 2- 3/442-14 '--- - - " - ^ ^ ~ " ' ' ' ' R ca.r r t:r- .- M:. 0 7'.'ait 21
~. __ . . _ _ _ . = ~ . .
4 for Specification-3,2,3-i Attach:nent- 1 : , 3.2.3 F,,(X,Y) 3 shall be limited by imposing the following
-; relationship:
FL (X, Y) i ( Fin (X, Y ) ] " l where: FL (X, Y) - the measured radial peak. [ F1, ( X, Y ) ] " - the maximum allowable radial peak as defined in. Core Operating Limits Report
-(COLR),
t J. 6 l
. _ . . - , .m __ _ _ _ . _ . . _ .,_ - ,, _ . , . , , - .. . , . , , . _ . . . . . . . , , _,
. . , ~ . - . . . . . . - - ~ . - . ~ . . - - -. ~ .. . - . ~ . . - - . - , _ . - -
for Specification 3.2,3 Attachment 2: ACTTON:-
' With Fa[(X, Y) - exceeding it a l!.mit :
- a. Within 2 hours, reduce the-allowable THERMAL POWER from RATED THERMAL POWER at least RRH % '" for each 1% that PL(X,Y) exceeds the limit, and
- b. Within 6 hours either:
- 1. Restore FL(X,Y) tc within the limit of Specification 3.2.3 for RATED THEBMALLPOWER,-or 2.- Reduce the Power Range Neutron-Flux-High Trip Setpoint in Table 2.2-1-at least RRH% for each 1% that TL(X,Y) exceecs
'that limit, and
-c.. Within 72 hours of initially being outside-the limit of Specification.3.2.3, either:
- 1. . Restore FL(X, Y) ' to 'within the limit of Specification 3.2. 3 - for RATED THERMAL POWER, or
- 2. Perform the following actions:
- (a) Reduce the OTAT iK term in Table 2,2-1 by f at 'least TRH '"
l for-each 1%-that FL(X, Y) exceeds the limit, and L (b) Verify through incore mapping thit FL(X,Y) is restored to within the limit for the reduced THERMAL POWER allowed by l_ -ACTION a', or reduce THERMAL POWEF to less than 5% of
. RATED THERMAL' POWER within the ne..t 2 hours.
l- [
'"-RRH-is the amount of THERMAL POWER reduction required .3. compensate for-each 1% that FL(X,Y) exceeds the limit of: Specification 3.2.3,.
provided -in the -COLR per Specification 6. 9.1.9. 42' TRH is the-amount of OTAT K 1 setpoint reduction required to compensate for each 1% that FL(X,Y) exceeds the limit of Specification 3.2.3, provided in the COLR per Specification 6.9.1.9.
(. 3, i ~ ' POWER DISTRIBUTION LIMITS 3/4.2.3 ""! ~' C'.' SIT ~ C0 NUCL OP FNTHALPY RISE WOT CHANNEL CACTOR - (y 6< , y) LIMITING C:NDITION FOR OPERATION 3.2.3 The ecmoination of inoicatea Reactor Coolant System (RCS) total flow rate ano R snaii be maintainea within the region of allowaole coeration , i specifieo in tne CORE OPERATING LIMITS REPORT (COLR) for four loop operation: Where: N F
)
- a. R=
p aTP (1. 0 MF g (1.0 - 0)] ' f Q r .Mi s ItE M CS wivH THERMAL POWER D' ,
- RATED THERMAL POWER i.
~
- c. = Measureo values of F cotainea by using the movaale intore detectors to obtain a power distribution map.
f The measurea values of F shall be used to calculate R since the figure l specified in the COLR includes penalties for unaetectea (
, feeowater venturi fouling of 0.1% and for measurement k uncertainties of 1.7% for flow and 4% for incore measurement ,
N
\, of F j, c N 3
d, = Tlie F limit at RATED THERMAL POWER (RTP) specifica in tne f I COLR, anc
) 3 MF = The power factor multiplier specified in the COLR.
-'(e . ._ .-
i APPLICABILITY: MODE 1. (Uni i !) t - { With the ecmoination of RCS total flow rate ano R outsice the region of acceptacle :ceration soecifiea in tne COLR: _I ,
- a. within 2 hours either:
'.. Restore tne comoination of RCS total flow rate and R
,.. to within the aoove limits, or PGLAG '\ l\
M rn 2. Recuce THERMAL POWER to less than 50% of RATED THERMAL POWER g g g- anc reauce the Power Range Neutron Flux - High Trip Setcoint L q kN to less tnan or eaual to 55% of RATED THERMAL POWER within the next 4 hours. /
/
j V ~ McGUIRE . NIT.~. ' :nc 2 3/442-14 Aac. a c.. N 0% Unit 1) uw.r i 22 ~ ~ . . 9:. CXUnit 21
1 i for Specification 3.2.3 i Attachment 1: . 3.2.3 ru (X,Y) shall be limited by imposing the following relationship TL(X,Y) $ (Ti,(X,Y)J W i where .rL(X,Y) - the measured radial peak. [ Fin (X,Y))* - the maximum allowable radial peak as defined in Core Operating Limits Report (COLR). t i o. f 9 k I l v ( F l [. l F
,-e -, - , + . , m--,..~,,en<-mem,.,--ne. ,.,,..--g ..,,-m.,,,e....nm.,-,,,a,n,.nnn,...m,,.,.,,m ..,Nn....,,,,.,,.xm-.w.w,n--.e ., s., , ,- -- --c'm -r.-, -c
_ _ . . _ _ - _ . _ _ . _ _ . _ _ . _ - . _ .__ _...___.m.m._ I I i for Specification 3.2.3 l 1 l Attac!unent 2: ACTIOf t : . l' ' With Fu (X, Y) exceeding its limit:
- a. W! thin 2 hours, reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH%"' for each 1% that Th(X,Y) exceeds the limit, and
- b. Within 6 hours eithert
- 1. Restore FL(X, Y) to within the limit of Specification 3.2.3 for KATED THERMAL POWER, or
- 2. in Reduce2.2-1 Table the Power at least RangeRRH%Neutron Flux-High f or each 1% thatTrip Fu (Setpoint X, Y) exceeds that = limit, and I
- c. Within 72 hours of initially being outside the limit of Specification 3.2.3, either:
i
- 1. Restore FL(X, Y) to within the limit of Spec 2ilcation 3.2.3 for- l RATED THERMAL POWER, or
- 2. Perform the following actions:
(a) Reduce the OTAT K term in Table 2.2-1 by at least TRH "' - 3 for each l% that FL(X,Y) . exceeds the limit, and
- (b) Verify through incore mapping that FL(X,Y) is restored to within the limit for:the reduced THERMAL POWER allowed by ;
ACTION a, or reduce THERMAL POWER to less than 5% of- , RATED THERMAL POWER within the next 2 hours.-
"' h5H is the amount-of THERMAL-POWER reduction required-to compensate for each.1% that . FL (X, Y) _ exceeds the limit of Specification 3.2.3, ,
provided in the COLR per Speciiication 6.9.1.9. d' TRH is the amount of OTAT M setpoint reduction required to compensate for each -1% that' FL:(X, Y) exceeds the limit of Specification . , 3.2.3, provided in'the COLR per Specification 6.9.1.9. e m*ge-*~p r'y &r- y - y* yWw y -
U@T' POWER DISTRIBUTION LIMITS LIMITING c NDITION F0k OPERATION ACTION: (Continued) pt#4 _ D. ! within 24 hours of initially being outside the above limits, verify ! through incore flux mapping and RCS total flow rate comparison that the ccmoination of R and RCS total flow rate dre restored to within i the above limits. or reduce THERMAL POWER to less than 5% of RATED i THERMAL POWER within the next 2 hours.
- l Q. ond/or c.2. .~
d./. Identify and correct the cause of/the out-of-limit condition prior to increasing THERMAL POWER abovel the recuced THERMAL POWER limit recuireo by ACTION (a.2. ano/or DJ above; subsequent POWER OPERATION may proceed provided that " - ~ 4aitd-- ;f R oou i..w.-w C00 %(x,y jg t:t;l 'l a - n; :r: demonstrated, through intore flux mappingpaad E: ;;t:! '!:u nt: :: ::r r% to be within the i
- -'er M :;;;;te?:
- , a .e , specified in the COLR prior to exceeding the following THERMAL POWER levels:
Lidt,}
- 1.
- 9 -d 50% of RATED THERMAL POWER, 3
- 2. r ":! 75% of RATED THERMAL POWER, and
- 3. Within 24 hours of attaining greater ti,an or equal to 95% of RATED THERMAL POWER.
SURVEILLANCE REOUIREMENTS Instet N M' 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 5 ~~-~ 4.2.4.4 ine companation of inoicatea ACS total flow rate ceterminea cy b process computer readings or digital voltmeter measurement and R shall be within the region of acceptable operation specified in the COLR: } M- a. Prior to operation above 75% of RATED THERMAL POWER af ter each fuel loaaing, and bed b. At least inte per 31 Ef fective Full- Power Days. Misc.her*4 4 5
~
4.2.3.3 ine inaicateo scs total riow rate snail be verified to ce witnin the region of acceptacle operation specified in the COLR at least once per 12 hours when tu most recently octained value of R obtainea per Specification 4.2.3.2, { is assumed to exist. 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATICH at-least once per 18 months. 4.2.3.5 The RCS total flow rate shall be determined by precision heat calance measurement at least once oer 18 months, b Delcke. (.ine.orporedAd in Sped [ic.ebion 4.2.5) McGUIRF, - UNITS 1 Ana 2 3/4A2-15 '
-- ----" M a i n
- H n " 11
- r : t ut. = U 't :T
I for Specification 4.2.3.2 Attachmeat 3: 4.2. 3.2 FL (X, Y) shall be evaluated to determine whether Fu(X, Y) is within its limit by:
- a. Using the movable incore detectors to obtain a power distribution map at any TliEFRAL PCWER greater than Si of RATED TilERMAL POWER.
for Specification 4.2.3.3 Attachment 4:
- b. Measuring FL(X, Y) according to the following schedule:
- 1. Upon reaching couilibrium conditions after exceeding by 10%
or more of RATED THERMAL POWER, the THERMAL POWER at which FL (:-:, Y ) was last determined"', or
- 2. At least once per 31 Effective Fu'.1 Power Days, or
- 3. At each time the OUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incore detector measurements.
"' During power escalation at the beginning of each cycle, THERMAL POWER may be increaued until a power level for extended operation has been achieved and a power distribution map obtained.
i i for Specification 4.2.3.2 Attachment 5:
- c. Perf orming t he following calculations:
- 1. For each location, calculate the 't margin to the maximum allowable design as follows:
PL(X,Y)
% Fu Margin - (1 - $ x 1001
[ Fi.i ( X , Y ) ] "" 11o additional uncertainties are required for FL (X, Y ) , because [ Fig (X, Y ) ]"" includes uncertainties.
- 2. Find the minimum matgin of all locations examined in 4.2.3.2.c.1 above. If any margin is less than zero, comply with the ACTIO!1 requirements of Specification 3.2.3 as if [Fia (X, Y) ]"" is t he same as [ FA, (X, Y) ] * .
- d. Extrapolating"' at least two measurements to 31 Effective Full Power Days beyond the most recent measurement and if:
FL(X,Y) (extrapolated) 2 [ FI, ( X , Y ) ] "" (e xt rapolated) Q (X,Y) f oxt rar>nl at ori) > $ (X,Y) ( FI, ( X , Y ) } "' ' (ext rapolated) [ F;, ( X , Y ) ) "" either of the following actions shall be taken:
- 1. FL(X,Y) shall be increased by 2 percent over that specified in 4.2.3.2.a, anu the calculations of 4.2.3.2.c repeated, or
- 2. A movable incore detector power distribution map shall be obtained, and the calculations of 4.2.3.2.c shall be performed no later than the time at which the margin in 4.2.3.2.c is extrapolated to be equal to zero.
- Extrapolation of for the initial flux map taken after reaching equilibrium conditions is not required since the initial flux map establishes the baseline measurement for future trending.
U Nl T
;0WER DISTRIBUTICN LIMITS 3/4.2.4 OUADRANT POWER TILT RATIO t.IMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceea 1.02.
APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWERk [] 11:[ l O i ) /L ACTION:
- a. With the QUADRANT POWER TILT RATIO dete" mined to exceed 1.02 but less than or equal to 1.09:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reducea to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Within 2 hours either:
a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for eacn 1*. of indicated QUADRANT POWER TILT RATIO in excess of 1.02and similarly reduce the Power Range Neutron Flux-High Trio Setpoints within the next 4 hours.
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours af ter exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setooints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and g 4 Identify ana correct th9 cause of .ne out-of-limit condition prior to increasing THERMAL POWE , suoseauent POWER OPERATION above 50% of RATED THERMALMmay proceed provider' that the QUADRANT Puh'.R TILT RATIO is verifiea within its limit at least s e sa - hot.c for 12 hours or until verified acceptable at 95%
or greater RATED THERMAL POWER, i l 'See Special Test Exception 3.10.2. l qg Not, a scoMe. udl cobbroSon cd the. ucore, debdors is compMed subsyt to r ue.Liu3 p cGUIRE - UNITS 1 ano 2 3/4 A2- 19 -- Hcnm.a t. 2 miii 1) ! .' annc=nt '!c.1: ("^it 2) h
f POWEP O!$TRIBUTICN L:MIT$
- LIMITING CONDITION 80R OPE 8ATICN a
ACTION: (Continueo)
- u. With the CUADRANT POWER TILT RATIO determinea to exceen 1.09 due to misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reouced to within its limit, or b) THERMAL POWER is reouced to less than 50% of RATED THERMAL POWER.
- 2. Reduce THERMAL POWER at least 3% f rom RATED THERMAL POWER for I 0L G, within
{ tach 1%30of minutes;- indicatec QUADRANT POWER TILT RA
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceecing the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next #
2 hours and reouce the Power Range Neutron Flux-High Trio Setpoints to less than or eaual to 55% of RATED THERMAL POWER within the next 4 hours; and 4 Identify ano correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsecuent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its ilmit at least once car hour for 12 hours or until verified acceptacle at 957 or greater RATED THERMAL POWER.
- c. With the QUADRANT POWER TILT RATIO determinea to exceed 1.09 due to causes- other than the misalignment of either a shutdown or control ron:
- 1. Calculate tne OUADRANT POWER TILT RATIO-at least once per nour t.ntil either:
a) The QUADRANT POWER TILT RATIO is reouced to within t its limit, or . L h) *WERMAL POWER is reoucea to less tnan 50% of RATED THERMAL POWER. 1 i l l: l . pub t 4
":GUIRE 'JNIT$ 1 e 3/4/,2-20 'm= r xt " 32 -Unu 1) D nc.t n. y (Unit 2) J
r
} .
l V DOWER O!STRIBUTION LIMITS LIMITING CON 0! TION FOR OPERATION ACTICN: (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours anc reouco the Power Range Neutron Flux-Hign Trip Setootnts to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
- 3. Identify and correct the cause of the out-of-limit conoition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceso provideo that ther QUADRANT POWER TILT RATIO is verif ten within its limit at least.
once per hour for 12 hours or until verified at 95% or greater-
~ RATED THERMAL POWER.
- d. % prem. .., 4 Jr4. A bh.= L o, Y ar e n e+ .pr .M I. .
1 I, - SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is OPERA 8LE, and
- b. Calculating the ratio at least once per 12 hours during staaoy-state coeration wnen the alarm is inopereole.
4.2.4.2 ~he QUADRANT POWER TILT RATIO shall be determinea to be within the limit wnen above 75% of RATED THERMAL POWER with one Power Range enannel inoceraale ey using the movanle incore catectors to confirm snat tne normalized synestrie :ower distribution, obtained from two sets of four symmetric thimole locations or a full core flux map. is consistant with the indicateo QUADRANT POWER TILT RATIO at least once per 12 hours. AIA> b l wc GUIRE - W IT$ 1uc 3/4tg-21 *rmt :. ::.U'i t M p ..- - w"a ;;Um :) l l _ _ _ _ _ _ - . .__ _.. __ _ _ __ _ --~
- _ - _ - _ _ . - . _ - _ - _ - _ _ . . - - . _ - .. . - ~ . - . _ - - - . __. -_
U cJ 1 T' i POWER OISTRIBUTION LIMITS 3/4.2.5 DNS PARAMETERS i LIMITING CONDITION FOR OPERATION 3.2.5 The following DN8 related parameters shall be maintainea within the limits shown on Table 3.2-1rA( vofn2}
- a. Reactor Coolant System T,,g, ene-
- b. Pressurizer Pressure.
~
- 0. ReactOP Ocold.n t 3 Sie m APPt.!CABILITY: MODE 1. (U ni 4 i) Ma i FM W M j ACTION:
iden4 Med n 3.2.E and b . &Ve a..With ei4anyhe ofrthe +eeve parameter:Aexceeoing its limit, restore the parameter to within its limit within 2 hours or reouce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours, e (Add Inced (D , anoched] SURVEILLANCE REQUIREMENTS 4,'24.1 e
' 2.P Each of the parameters of Table 3.2-1 shall be measured by averaging the indications (meter or computer) of the operable channels and verified to be within their limits at least once per 12 hours.
hdci 1nwr-4 @ , athched
~
McGUIRE - u"!T: 1 == 2 3/.%2-22 ' enxx: ":M ("t Q frran .; .i ;.eM . 6 -r up iT , t gaf 2, l.
l l l \ for Specification 4.2,5 lt! SERT 1: System total ilow tate and
- b. With the combination of Reactor Coolantr egion of restrict ed operation specified TilERMAL POWER within the on Figure 3.2.1, within 6 hours reduce the Power Fange ticutron Flux-High Trip Setpoint to below the nominal set poi nt by the same 1.
amount (% RTP) as the power reduction required by Piqure 3.2
- c. With the comlJination of RCS total flow rate and TilEIO4AL POWER within the region of prohibited operation specified on Figure 3.21:
- 1. Within 2 hours either; flow rate and THERMAL a) Restore POWERthe combination to within of RCS the region total of permissible operation, Reactor Coolant System total b) Restore the combination of flow rate and TilERMAL POWER to within the r egion of or above, restricted operat ion and ccmply with action b.
c) Reduce TilERMAL POWER to less than 50% of RATED TilERMAL POWER and reduce the Power Range tieutron Flux - High Trip Setpoint to lens than or equal to 55% of RATED TilERMAL POWER within the next 4 hours.
- 2. Within 24 hours of initially being within the region of that verify prohibited operation specified in Figure 3.2-1, flow rate are the combination of TIIERM.AL POWER and RCS total restricted restored to within the regions of permiscible or operation, or reduce TilERMAL POWER to less than 51. of RATED THERMAL POWER within the next 2 hourc.
ItJSERT 2 : The RCS total flow rate indicators shall be subjected to a 4.2.5.2 CilAtitJEL CALIBRATIC11 at least once per 18 montho.
~ shall be determined by precision heat
~4.2.5.3 The RCS total flow rate balance measurement at least once per 18 montho.
I
LL M iT I TABLE 3,2-1CA. 28 PARAMETERS v OPERABLE PARAHETE9 IN0! CATION jHANNELS t.!MITS* ) Indicatta Reactor Coolant System i'V9 rneter 4 <590.5'F h meter 3 [590.2'F / comouter 4 < 591. 0* F computer 3 5590.8'F Indicated Pressurizar Pressure'" meter 4 32226.5 psig meter 3 3,2229.8 psig computer 4 3,2221.7 psig computer 3 3,2224.2 psig A Reacle Ccolod SgskmTohl Flow Rh.+e. Figure. 3.2 - 1 ( L
.' i 4
\
/
s 4
' Limits appl 1 cable during four-loco operation.
**t,1mits not cro!4 cable during either a THERMAL POWER rame in excess of 5% of RATED THERA PufER per minute or a THERMAL POWER step in excess of 10%
RATED THERMAL POWER. uc GUIRE - ;N:T" I d 2- 3/442-23 Jzr n 2 M " '""" ' b
?- - ..- .~.M G n 2 )
w s.r3 1 J A>4 2
....--~.mJ-.,--.m . . . . , , _ . - . ~ . . - ~ . , - . , . _ - - - - , - - , - . , . - , . , _ . , - . . . .--,1-.. .-.--+--.-..-..r-;
Vt>IT 1 Figure 3.2 - 1. Reactor Cootent System Total Flow Rate Vereue Rated Thermet Power . Four Loops in Operation 388850 Permissible A pensity of 0.1% for undetected feeowater Operation ventun fouling and a measutomont uncertainty *" cf 1.7% for flow are included in this figure. . (98.385000) 3850]O - ---------~~~------------~~~~~~~-~~~~~~~ (96,381150) E n. 3 M1150 - Restncted
- , Opomtor.
Region . ! > (94.377300) C 377300 - E a Prohibited g - Operanon
. , (92.373450) Regen , 373450 o
U ,
.o (90,369600)
{ c: - 369600 365750 361900 , 100 102 88 90 92 94 96 90 86 Fraction of Reted Thermel Power McGuire - Unit $1 4 4 1.
UNIT 3 IAULL 3_3-Ic( N c3 C 111 ACION INIP SYSIIH lilSilillMitil Allute M m HitilHi!H EllAtitit t S CliANt4LI S APPtICAUt[
. 101 At tiu. HODES AC110ri TO TRIP OPERABIE c of CilAtitiEI S $ 111tici l0t4A1 titill 1, 2 1 u Hasiual iteac t.as erip i
2 1 2 2 3 * , 4" , 5* 10 g 1. 1
. 2 1, 2 2 l 3
a Power Range, ficutrusi i lux - liigh 4 2 2. y Setpoint 2 3 1,,,, 2 2 l low 4 5etpoint 1, 2 2 l 2 3 4
- 3. Pouer llatige , tietstroni i lis< -
Aligh Positis i Rate ~~ w O d.c}C. _ __. - 2 1 1, 2 / ls s 4
- 1. ruwer li.uiuc , tien t i nii ilo=,
ku w_ liigh liegative Rate I 2 l ##', 2 3
?
S. Intermeiliate Hange,tieutron fI.<
- 6. Source Hange, taeution ilux 1 2 2 4 2 5* 10
- a. Startup 1 2 3*, 4*,
2
- b. Shutdown 0 1 3, 4, and 5 5 2
- c. Sinildown h
/. Overtemperature al 1, 2 6 l 2 3
- 2 a 4 (aa)
a' Iour ioop Operatiosi (**) (aa) (aa) lhree loop Operation (**) lf J. 22 -lil
.,+
<J
UNI I L lABLE 3.3-2 %
- c n ESPotist ilMls O _h_1 AC.1_0_H. .I HIP $) $ llit. .i.l.151Hu!1 Ell.I A. T 10N. R
- n m
ill SPOf45i Ilill [z l Hut I lut4At tite l I I4 A 3, ! !1 ino.i i I;e... t . .: Irip u u 5 second (I) P .wer ILuige . tient : un i lu. a o Ndttge , lleutt'ull l lal A ,
.) l'Uwe l' ti. A . " liigh Positive Rate ~ ~ '
blglo . .-- - ._ _. h.wer Hasige , ficistron I linx, 2< [4 liigh lieg,itive Rate __ _- 105 second (1)
~ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _
l fi . A S I nt ermedi at e it nge, ficut run i lin l w s .n i lo-ti A a t. tnori e itinge rJeut i B 2 10.0 seconds (1)(2)JWQ [
.L / loestemperatuie al 4 ,
-141 0 ,etonds (1)(2)M n INc [..wei al Pi cssur i zer l'i essure - t ow 120 seconds U
102 seconds - > las P essur i zer l'a cssur e- -lisyli i, fp NA 2 18 l'acssuiize: W.el e s ievel -liigli g e> CP Response time of the neutron flux signal portion f (1) fleutron detectors are exess:1 L f ruir .sponse time testing. in cnannel of the channel shall lie meass:.ed in om detector output or input of first electronic component 7 M (<1 T r.e ; 10.0 secunal response t ime incluiles a 6.5 second delay for the RIDS mounted in thermowells. q
- ,m;.. ce 1:-- -
_,j;j, a 7e ic; cog m,;; ca;y 2fts.p_g;u n;;; g, g.,,_ ,,,3 a g g e y g 7 -4 ~ : !!m _9 in n .. .. - -.
. u .. . , . . . , J a -
IAltl l 4. 3-I R - N g klACIUM litil' SYSil!! It4SIRtI!11t41 AII0t1 StiRVlill Ar4CE REQUIRLHEt415 "o IRIP m
- AtlAIOG ACIIIAllt4G H0 DES 10R c CilAlitlEl DEVICE Wilitil 5 CllAtitill tilAlit4E L OPERA 110t4Al DPERAT10t4Al. ACIUA110ft SilRVElll At4LL U litr4Cliot4A1 titill CilECK cal lBRAII0tl TEST TEST 10GIC IESI Is REQUIRED-r
- 1. fl u n e.s i lle.u tur t rip ti A . ti A . fl . A . R (11) 14. A . I , 2 , 3 * , 4 ' , *. ' l ru 2. Power Range, tient un i lux liigli Setpoint S D(2, 4), M ti . A . fl . A . 1, 2 i M(3, 4),
Q(4, 6), R(4, 5) low Setpoint 5 R(4) H I4. A. tiA. 1,,,, 2 LJ ti. A. ii . A . 1, 2 g 3. Powe- Range, tietit e nti i lii , it. A. l((4 ) H w liigh Positive Rate ' 3 Dalci,e __ - _ _ _ . _ _ . _ _ ._ - e ( 4. Power Range, tieutron flux, liigh flegat ive Raler it . A . !!(4 ) H ti . A . 14. A . 1, 2
]
C _
- 5. Intermeiliaie Range, S R(4, $) $/U(1),H ti. A. ft. A. I ,.'
fieutron flux
, j, b. Suurte Itange, fleutsun Ilum S 14(4, 5) 5/tJ( 1 ),tt( 9) fl . A . ft. A. 2 , 3, 4, 5 j a a i El /. Overtempeiature il S R H ti . A . f4 A. I, 2 El t
{ u. Uverpowe- al S R H 14. A. ft. A. 1, 2 !
- 9. Pr essuri zer Pressni e--l ow H ft. A. ft. A.
lI S H 1 m . Pressui i 4 er Prest.us e--lingle H ti . A . ii . A . 1, 2 l y 10. S H 4:> N7 . Il Ps i- sor t ii 4 W.s t e s l e ve l - - li t e;h . R 11 f4 A ti A I i ' 12 los 1 ca, t ., tou t u.t i I ..w 1: !! 14 A 14 A . I l i l l 4
, Mver .[ INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT!0N LIMITING CONDITION FOR OPERATION 3.3.2 The Cngineered Safety Features Actuation System (ESFAS) Instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-43and with RESPONSE TIMES as shown in Table 3.3-EA (t/,vrrJL\
'+(U4nT i) J APPLICABILITY: As shown in Table 3.3-3. ( (( ol t i o.3)g)
ACTION:
- a. With an ESFAS Instrumentation channel or interlock Trip Setpoint less conservative than the valun shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the Trip Setpoint adjusted consistent with the Trip Setpoint value,
- b. With an ESFAS Instrumentation channel or interlock inoperaole, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS Instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2, 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shale include at least one train such that both trains are tested at least unce per 36 months and one channel per function such that all channels are tested at least once per H times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" column of Table 3.3-3. McGUIRE - UNITS 1 and 2 3/4A3-15 ___,___._m_____
l lABLE 3.3-4 m 1 Et4GitifERED SAFETY FEAlliRES AtitlATIO!1 SYSTEH It1STRtIMENIATI0tl TRIP SETPOINTS D1 IRIP SEIP0lNT Alt 0WABLE VALUES >
- I titicl10NAl 18111 C
z 1. Safety Injection, Reactor Irip, fee <! water Isolation, Component Cooling d Water, Stari Diesel Generators, and g fluclear Seivice Water.
? fl. A. N.A.
m
- a. Manuai Initialion ti. A. fl. A.
- b. Automatic Actuation Logic and Actuation Relays Cunt.. .unent l>ressure--liigh 1 1.1 psig $ 1.2 psig t
1 1845 psig 1 1835 psig it . Piess...izer Piessure--low-tow psi 1 psig Steam iine Picssure - low e. 1 [TTS . SS to an ! Conta i simesi. Spray
- a. tionnai initiation II A- li' ~
H.A. fl. A.
- b. Autoin..Lic Actuation Logic and Actuation Relays Contaisiment Pi essure--liigh-liigh 1 2.9 psig i 3.0 psig
, c. bs M
UM 7 1 TABLE 3.3-4a(Continued) #N E #$ deJf7 N ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETFOINTS g x W. IRIP SETPOINT ALLOWABLE VALUES
- FUNCTIONAL UNIT C
{ 3. Containment Isolation w
- a. Phase "A" Isolation N.A. N.A.
- 1) Manual Initiation I
N.A. N.A. 2)~ Automatic Actuation Logic and Actuation Relays Safety Injection See item 1. above for all Safety Injection Trip Setpoints 3) and Allowable Values w b. Phase "B" Isolation 2 H.A. y 1) Manual Initiation N.A. l N.A. N.A.
- 2) Automatic Actuation Logic and Actuation Relays Containment Pressure--liigh-High 1 2.9 psig i 3.0 psig 3)
- c. Purge and Exhaust Isolation N.A. N.A.
l 1) Manual Initiation I N.A. N.A.
- 2) Automatic Actuation Logic and Actuation Relays Safety Injection See Item 1. above for all Safety Injection Trip Setpoints
- 3) and Allowable Values l
,, Ld
M TABtE 3.3-4a(Continued) c1 5 LliGINEERED SAFETY FEATilRES ACillATI0!i SYSTEH IllSTRtittEllIATI0tt TRIP SETPOIllTS M f ittlC I 1011A1 IINIT TRIP SETPOINT AttOWABtE VAtilES s y, 4. Steani 1 isie Isolationi e p a. Manuai Illili aliof) II. A. ti. A. ro
- b. Automatic Actualion Logic fl. A. tL A.
and Actuation Relays
- c. Contais. ment Pressure--High-liigh 1 2.9 psig i 3.0 psig
- d. Negative Steam Iine i 100 psi with a i 120 psi with a m Pressure Rate - liigh rate / lag function rate / lag function 1 time constant time constant t __
> 50 seconds > 50 seconds 4
N
- _1
- e. Steam Line Pressure - tow > g psig > 365'psig 776 755
- 5. Turbine Trip and feedwater Isolation
- a. Automat oc Actuat ion logic N.A. ft. A.
and Actuation Relays Eiii b. Steam Generatoi Water ievel-- < 82% of narrow range < 83% of riarrow v'asige hk High-liigh (P-14) Instrument span each steam Instrument span each steam llg generator generator [ c. Doghouse Water Level-liigh 12" 13" jrji (Feedsater Isolation only) C)*= j 6. Containment Pressure Control System SE!? Start Permi_,3 ve/leamination 0. 3 s~ $P/T < 0.4 PSIG U.25 SP/l O 45 I";In (SP/I) ~ (~ ~ Q 7- ~ e L
4//I 1. (E5AhE A5(bJtT L TABLE 3.3-4M Continued) N E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E 1 l FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES
] 7. Auxiliary feedwater
- a. Manual Initlation H. A. N.A.
@ b. Automatic Actuation Logic N.A. N.A.
[ and Actuation Relays
- c. Steam Generator Water Level--Low-tow
- 1) Start Motor-Driven Pumps > 12% of span from 0 to > 11% of span from 0 to 30% of RATED THERMAL POWER, 30% of RATED TliLRMAL POWER, increasing linearly to increasing linearly to M > 40.0% of span at 100% > 39.0% of span at 100%
if RATED THERMAL POWER. if RATED TiiERMAL POWER. { E
- 2) Start Turbine-Driven Pumps >12". of span from 0 to > 11% of span from 0 to 30% of RATED TitERMAL POWER, 30% of RAILD iltERMAL POWER, increasing linearly to increasing linearly to
> 40.0% of span at 100% > 39.0% of span at 100%
f RATED TifERNAL POWER. 5f RATED illERMAL POWER.
- d. Auxiliary feedwater > 2 psig -> 1 psig EE Suction Pressure - Low
@@ (Suction Supply Automatic Realignment)
EE e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints 2z Start Motor-Driven Pumps and Allowable Values oo
> 3200 volts I
- f. Station Blackout - Start 3464 1 173 volts with a M Motor-Driven Pumps and 8.5 1 0.5 second time E2 Turbine-Driven Pump delay
- 3. 3.
er (Note 1) e l m ._. g. Trip of Main feedwater Pumps - N.A. N.A. Start Motor-Driven Pumps - i
$1l ' W L' (a/m15 m (_;,ygt 2_
TABLE 3.3-43 (Continued} N E2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E m
- TRIP SETPOINT All0WABLE VALUES FUNCTIONAL UNIT "z
- 8. Automatic Switchover to Recirculation
{*
"" RWST Level 3 90 inches 3 80 inches o
" 9. Loss of Power
~
4 kV Emergency Bus Undervoltage- 3464 1 173 vo!ts with a 3 3200 volts l i Grid Degraded Voltage 8.5 1 0.5 second time delay
- 10. Engineered Safety features Actuation System Interlocks Pressurizer Pressure, P-il < 1955 psig < 1965 psig t', a.
ib, > 551 F
- b. T avg, P-12 -> 553*F -
n3 e N.A. N.A.
- c. Reactor Trip, P-4 Steam Generator Level, P-14 See Item 5. above for all Trip Setpoints and Allowable d.
Values. EE The turbine driven pump will not start on a blackout signal coincident with a safety injection signal.
@@ Note 1:
EE Ba
.sn O.
as
- 3. "
ee L-_ -- - -
f.]l I ABLE :.2-Ect ENGINEERED SAFETY FEATURES RESPONSE TIMES INITI ATING 5IGNAL AND FUNCT CN RESPONSE T ME IN SECCNE:
- 1. Manual
- a. Safety injection (ECCS) N.A.
- b. Containment Spray N.A.
- c. Containment Isolation Phase "A" Isolation N.A.
Phase "B" Isolation N.A. Purge and Exhaust Isolation N. A.
- d. Steam Line Isolation N.A.
- e. Feeowater Isolation N.A.
- f. Auxiliary Feeowater N. A.
- g. Nuclear Service Water N. A.
- h. Comoonent Cooling Water N.A.
- i. Reactor Trip (from SI) N.A.
- j. Start Diesel Generators N:A. _ _ _ - .
- 2. Containment Pressure-High
- a. Safety Injection (ECCS) i 27(1)
- b. Reactor Trip (f rom SI) i2
- c. Feeowater Isolation </ 12.
)
- d. Containment Isolation-Phase " A" i 18E )/28
- e. Containment Purge and Exnaust Isolation _
4 i f Auxiliary Feeowater( } N.A.
- g. Nuclear Service Water < 65(3)/76(4)
- h. Component Cooling Water 1 65(3)/76I) 1 Start Diesel Generators i 11 PdcGUIRE - UNIT 51 & 2 2/4/j3-30 1 emenW4o,102 (Ur4 +
% nc :ent *-!ca^ (Ur11 .
2/10/90
IU -l TABLE I.1-5N Continueo) ENGINEEEED SAFET :EATURES RESPONSE T"1ES "itTIATING SIGNAL AND FUNCTICN RESPONSE 'IME IN SEC:NCS
- 2. Pressurizer Pressure-tow-Lew
- a. Safety Injection (ECCS) 27(1)/12(3) i
- b. Reactor Trip (from SI) <2
~
- c. Feeawater Isolation < / L 2.
- d. Containment Isolation-Phase ~'A"(2) i 18(3)/28(4}
- e. Containment Purge and Exhaust Isolation i4
- f. Auxiliary Feedwater(5) l N.A.
- g. Nuclear Service Water System < 76(1)/65(3)
- h. Comoonent Cooling Water
_ 76(1)/65(3)
- 1. Start Diesel Generators 11 2
Steam Line Pressure-tew
- a. Safety Injection (ECCS) 1 12(3)/22(4)
- b. Reactor Trip (from SI) <2
~
- c. Feedwater Isolation < # 12.
- d. Containment Isolation-Phase 'A"(2) 18(3)/28(4)
- e. Containment Purge and Exhaust Isolation 14 l
- f. Auxiliary Feeowater(U) N.A.
- g. Nuclear Service aater <
_ 65(3)/76(4)
- h. Steam Line Isolation < /'l o
- i. Comoonent Cooling Water 6553)/76(4)
- j. Start Diesel Generators _ 11
- 5. Containment Pressure-Hich-Hioh
- a. Containment Soray < 45 l b. Containment Isolat. ion-Phase '" .
- 4. A.
- c. Steam Line Isolation _ .7 t o S. Steam Generator Water Level-Hion--ich
- a. Turbine Trio 1. A .
- b. Feeowater Isolation i f ( 2, l
l l l l ucGUIRE - UNITS 1 & 2 3/4A3-31 -Menoment-Nor4G60ni t ' ' 4 enoment-No.04 (Uni- 2', __ _ _2/10/90..
1 I Y A] W $ l
~13LE : . 2-in( Cent uec )
ENGINEERE: IWET' CEATURE! ;E5PONSE ~:9ES 1 l1ITIATING !?GNAL AND FUNCT::N :E!PONSE "vE *N 1E::ND5
~
l Steam Generator dater _e<ei _ :w t. c w
- a. Motor-criven Auxii' try Feeawater Dumos ; 60 D. Turcine ariven Auxi 'ary Feeowater Pumps 1 60
- 5. Necative 5 team Line Pressure Rate - Hion Steam Line Isolaticn 1 7 to
- 3. Start Permissive Containment Pressure Contro1 Sy5 tem N. A.
' 0.
. Terminati0n Containment Pressure C;ntrol System N.A.
*1.
. Auxiliarv ~eedwater Sucti:n Pressure - Lew Auxiliary Feeawater umos (Suction Sucaly Automatic Realignment) 1 13
*2.
. RWST Level Autcmatic Switchover to Recirculation j 60
- 13. Station Blackout
- a. Start u otor-0 riven 4xt i f ary Feecwater Pumps y 60
- c. Start Turoine-Oriven Aux 1iiary Feeowater Dumo (6) ; 60
' 4.
. Trio of Main Feecwater ;;mos Start .*otor-0 riven Auxiliary.
Feecwater Pumps j 60
'5.
. Loss of Ocwer a ei E ercency Eus cermitsce- ~
Gric Cegraced' Voltage
":GUIRE - tNIT3 ~. ano 2 2/463-32 ' cncment ';c. ~; s?'t 2
'::ncment 40.97 '"a4i 'J -
I HOT STANDBY , 1 LIMITING CONDITION FOR OPERATION ' 3.4.1.2 At least three of the reactor coolant loops listed below shall be OPERABLE and a lees, we of these cesctec cecient ic;p: ;hal Le in operation:"
- 3. Reacter Coolant Loop A and its associated steam generator and reactor coolant pump,
- b. Reactor Coolant Loop B and its associated steam generator and rdactor coolant pump,
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and
- d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.
APPLICABILITY: MODE 3 t ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTOOWN within the next 12 hours.
b. Lt4s h .0 4. A % rs d eed W.r WithgMy Onc reactor coolant loopsin operation, restorept 3y'rd - . , .
-two loops to operation within 72 hours or open the Reactor Trip '
System breakers. ,
- c. With no reactor coolant loop in operation, susp.end all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.
SURVEILLANCE REOUIREMENTS >. M rle-
"4.4.1.2.1 At least the above required reactor coolant pumps, if not in '
operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. I a 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 12% at least once per 12 hours. 2 -04, obwc. ree'<e.d 4.4.1.2./ At least % reactor coolant loops shall be verified in operation l and circulating reactor coolant at least once per 12 hours.
^All reactor coolant pumps may be de-energized for up to I hour provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core nutlet temperature is maintained at least 10*F below saturation temperature. McGUIRE - UNITS 1 and 2 3/4 4-2 Amendment No. '(Unit 1) Amendment No. (Unit 2)
~ ,
- ~
dhI giSTANDBY LIMITING CCNDITION FOR OPERATION 3.4.1.2 At least three of the reactor coolant loops listed below snall be l lMS: . . n 6 . ., v v. ^ ha n a n
. d..m ' ,n : h d ' h i n . J operation:' '
- a. Reactor Coolant Loop A and its associated steam generator-ano reactor coolant pump, l
- b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,
- c. Reactor Coolant Loop C and 1ts associated steam generator and reactor coolant. pump, and
- d. Reactor Coolant Loop D and its associated steam generator ano ,
reactor coolant pump. , APPLICABILI~Y: MODE 3 (L{ n i f / o n I tg. ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, restore the reauired loops to OPERABLE status within 72 hours or be-in HOT SHUTDOWN within the next 12 hours. I Laas Withi: $ e%n 4 r ,h..W 4 G,. restore n .resy.M s reg.A salreactor coolant loopsin operation, b.
>tr loops to operation within 72 hours or open the Reactor Trip System breakers.
- c. With no reactor coolant loop in operation, suspend all operations involving a reouction- in boron concentration of the-Reactor Coolant System and immediately initiate corrective action to return the reouired reactor coolant loop to operation.
SURVEILLANCE REOUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not -in operation, snall be determined OPERABLE once per 7 days by verifying correct breaker alignments and. indicated power availability. 4.4,1.2.2 ~1e requireo steam generators snall be oetermineo OPERABLE cy l verifying secondary. side water level to be greater than or equal to 12*. at least once cer 12 hours.- nree. 4.4.1.2.3 At least Jutr reactor coolant locos shall be verified in operation l and circulating reactor coolant at least once per 12 hours.
*All reactor coolant pumps may be de-energized for uo to 1 hour provicea:
(1) no operations lare permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintaineo at lear.t 10'F below saturation temperature. l l McGUIRE - UNITS 1 and 2- 3/4/Y-2 4menement h C (Unit li
-Amencaent Nc. 4 (Unit 2) - _ . . . - ~ _
Y/L]I $ REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve snall be OPERABLE with a lift setting of 2485 psig +.A %.*
+ 3 %,- 2 %
APPLICABILITY: MODES 4 and 5. ( u p ff f O $ [ ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. SURVEILLANCE REOUIREMENTS _ 4.4.2.1 No additional requirements other than those reouired by Specification 4.0,5. l l
*The lift setting pressure snall correspond to amoient conditions of the valve at nominal operating temperature and pressure.
l s prJ/ L Ms@UIRE - UNITS 1 ane 2 3/4/4-7
A)IT 1 REACTOR COOLANT SYSTEM OPERATING _L_IMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig 1 %^?y'2
+3' %
APPLICABILITY: MODES 1, 2, and 3 (0JIT.1)
/
i ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. i SURVEILLANCE REQUIREMENTS ~ 4.4.2.2 No additional requirements other than those required by Specification 4,0.5. l l l l
"The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
i McGUIRE - UNITS 1 and 2 3/4A4-8 1
I 3/4.5 E'iE:tGENCY CORE COOLING SYSTEMS 3/4.5.1 y ~4ULATORS COLD LEGj JTION LIMITI_NG CONDITION FOR OPERATICN 3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with,
- a. The isolation valve open,
- b. A contained borated water volume of between 6870 and 7342 gallons,
- c. A boron concentration of between 1900 and 2100 ppm,
- d. A nitrogen cover pressure of between 585 and 639 psig, ano
- e. A water level and pressure channel CPERABLE.
APPLICABILITY: MODES 1, 2, and 3* ACTICN:
- a. With one accumulator inoperaole, except as a result of a closed isolation valve or boron concentration less than 1900 ppm, restore the inoperable accumulator to CPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours ana recuce tr"" i :m Reocier pressure to less than 1000 psig within the f ollowing 6 hours. Coob-t syst u
- b. With ene accumulator inoperable due to the isolation valve being closed, either 1mmeciately open the isolation valve or be in at least HOT STANDBY within 6 hours and reduce ;-?: n . 6 fpressure to l?ss than 1000 psig within the following 6 hours. R ec ct.o r Cooto,4 Sygic,s
- c. With one accumulator inoperaole due to boron concentrr tion less than 1900 ppm and:
- 1) The volume weighted average boron concentration of the m M etting accumulators 1900 ppm or greater, restore the in-operable accumulator to OPERABLE status within 24 hours of the low baron aetermination or be in at least HOT STANDBY within the next 5 hours ana reduce prm:_. _s pressure to less than 1000 osig within the following 6 houh fecdor Coolat Syctc.
- 2) The volume weignted average ocron concentration of the tvec lic.iting accumulators less than 1900 ppm but greater tnan 1600J56tf ppm, restore the inocerable accumulator to OPERABLE status or return the volume weigntea average boron concentration of the three limiting accumulators to greater than 1900 ppm ano enter ACTION c.1 within 6 hours cf the low ooron determination Jr ce in HOT STANDBY within the next 6 hours ana reduce .g-er _. :"N Rea g pressure to less than 1000 psig within tne following 6 nours. coog,n t ytt w
'R +m i;m pressure above 1000 psig.
Reacher Coo \o.:t f.yh
.3 p))> L McGUIRE - UNITE 1 prd 2 3/4 5-1 h n-a n * ": ; i "i t 1) kvwxvwam
EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) soo
- 3) The volume weighted averagg oron concentration of the shr4a li*iting- accumulators 15Wp m or less, return the volume weighted average Doron concentration of the three limiting accumulator to greater than 35CKr ppm and enter ACTION c.2 within 1 hour of the low boron determination or be in HOT STANOBY wi;hin the next 6 hours and reduce p M4+r pres-sure to less than 1000 psig within the foll owing 6 hours.
Re.ctbr Coobt Shshm SURVEILLANCE REOU PEMENTS - 4.5.1.1.1 Eatn cold leg iniection accumulator shall be demonstrated OPERABLEL
- a. At least once per 12 hours by:
- 1) V;rity'.ig .he containeo borated water volume and nitrogen c.eer pressure in the tanks, and
- 2) Verifyirg that each cold leg injection accumulator isolation 51 . is open.
- b. At least .,ce per 31 days ana within 6 hours af ter each solution volume increase of greater than or equal to l% of tank vctume not resulting from normal makeuo by verifying the boron concentration of the accumulator solution; [
- c. At least once per 31 cays wnen the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected; and P Y
- c. At least once per 18 months oy verifying proper operation of the p power disconnect circuit. 4 4.5.1.1.2 Each cold leg injection accumulator water level and pressure channel snall be demonstrated OPERABLE:
- a. At least once per 31 days by the performance of an ANALOG CHANNEL
, OPERATIONAL TEST, ana
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
I J A thD E AcGUIRE uNITF 1 A w J 3/4 5-2 3
- n"mont Nn ( U '"+-rf
%cqd= nt_Ns. 7 (Unic Q
1 IURVEIll.ANCE REQUIREMENT 5 iC:ntinueal
- 2) A visual inspection of tne containment s.mo anc verifying that the suosystem suction inlets are not restrictea oy cebris ana tnat tne sumo c:moonents (trash racks, screens, etc. ) snow no evicence of structural distress or abncrmal corrosion,
- e. At least once per 13 months, during shutdown. ty:
- 1) Verifying tnat each automatic valve in tne flow path actuates to its correct :osition on Safety injecti:n actuation and automatic switenover to Containment 5 mo Recirculation test signals, and
- 2) Verifying inat eacn of the following pumos start automatically uoon receipt of a Safety Injection actuation test signal:
a) :entrifugal charging pumo, b) Iafety Injection pumo, and c) RHR pump. O r. ex ver4< vie 9 :=>t eeca er tae r 11e 4e9 pemo,=eveieas t"e imaicetee differential cressure when testec pursuant to 50ecification 4.0.5: l 2333
- 1) Centrifugal cnarging pumo > .::::-- o s i d ,
Msq
- 2) Safety Injection oumo > MCe.osic, anc
~ \68
- 3) RHR cumo > --s osu.
- g. By verifying the correct oosition of eacn e metrical and/or mecnanicai :csition stop for tne following EC:i tnrottle valves:
- 1) Witnin a hours following C0moletion of f acn alve stroking operation or maintenance on the valve wnen tne ECC5 suosystems are recuirea t: ce OPERABLE, anc O
a c0UIRE . NIT 3 . na . ;c4 5-7 -r.enament No. 97 (Unit 1) menament No. 79 (Unit 2)
4 I EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2) At least once per 18 months.
Boron Injection Safety Injection Throttle Valves Throttle Valves Valve Number Valve Number NI-480 NI-488 NI-481 NI-489 NI-482 NI-490 NI-483 NI-491
- h. By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
- 1) For centrifugal r.harging pump lines, with a single pump running:
a) Tbs sum of the injection line flow rates, excluding the h ghest flow rate, is greater than or equal to 443 gpm, ana 33 5 b) The total pamp flow rate is less than or equal to 565 gpm.
- 2) For Safety Injection pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 462 gpm, and 405 b) The total pump flow rate is less than or equal to 660 gpm.
- 3) For RHR pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3975 gpm.
Amendment No. (Unit 2) McGUIRE - UNITS 1 and 2 3/4 5-8 Amendment No. (Unit 1) 4/13/83
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION All main steam line Code safety valves associated with each steam \ 3.7.1.1 generator shall be OPERABLE with lif t settings as specified in Table 3.7-3G. (drM1) APPLICABILITY: MODES 1,2,and3.(OuTTI. ACTION:
- a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE stat.us or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be-in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With three reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. l McGUIRE - UNITS 1 and 2 3/4A7-1
/
h W N
- N N m . ,C C
N, N, N, y .. . . . U e e O
- N O O e w %. - - Q C C m m N N @ J LO
..c g - - - -
g O C
.a to G3 9
m O a
+1 3 3
w
.3 .3 .3 < .
g 1 a + m m a a O '*
. % % 3. L L e O ***
O D O D C a >= O
> 5 CD O N N N
.o.
g W - - N N g w w r== - - - -
~
m C m w a s W
- U
> -I e a a
< C m > w m >= n
>"" is W W 4 A w m < Q B .a 4 m Z C
- O J Q' a
= a N m e A D $
<C C L w Q > > > > > L .
n a m m m A m oe W u6
=
-m
-m o N m e.
O m - c. O O m - - - 2 5. o > > > > > ou a m m A A A uc am A AW
- 2 5 e o e s m : ;
a - - - - o ,u o > A A A A m =w
. _~
an
.a o v .a mi #
w < a rc O - N M 4 a z O c. o N N N N N . . _ _ .c..
.- a
= = > > > > > -e A A A n A g w ww
>, c c.
* *C e .
M
> - N i .S 5 u>6 2 McGUIRE - UNITE 1 aw# 3/4A7-3
l
- LANT SYSTEMS ///
MAIN STEAM LINE ISOLATION VALVE 5 LIMITING CCNDITION FOR OPERATI:4 3.7.1.4 Each main steam line salation valve (MSLIV) shall be OPERABLE, APPLICABILITY: MODES 1, 2, ano 3. [()g ff 1 O jt.Y ACTION: MODE 1 - With ore MSLIV inoperaole but open, POWER OPERATION may continue provided the inoperaole valve is restored to OPERABLE status within 4 hours; otherwise, reouce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours. MODES 2 - With one MSLIV inopersole, subseouent operation in MODE 2 or 3 may sno 3 proceea provided:
- a. The isolation vaive is maintained closed, anc
- b. The provisions :1 Soecification 3.0.4 are not apolicaole.
Otherwise, be in HOT ITANDBY within the next 6 hours and in HOT SHUTDOWN within the f0llowing 6 hours. SURVEILLANCE REOUIREMENTS a.7.1.4 Eacn MSLIV-snall be ce-enstrated OPERABLE by verifying full closure witnin /'seconcs wnen testea cursuant to Soecification 4.0.5. B 6 ffdh 2. c . . '. E E .14 IIJ :. A # 1 4f-d
ADMINISTRATIVE CONTROLS i CORE OPERATING LIMITS REPORT 7 ( 6.9.1.9 Core operating limits snall be established and documented in the i CORE OPERATING LIMITS REPORT before each reload cycle r any remaining part () of a reload cycle for the f9110 wing:
- 1. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm (
surveillance limit for Specification 3/4.1.1.3, )
- 2. Shutdown Bank Insertion Limit for Specification 3/4. ( ,
i
- 3. Control Bank Insertion Limits for Specification .'/4.1...o.
4 Axial Flux Difference limits, target band, and APLND*foc
" ( )
Specification 3/4.2.1, [
* ** N
- 5. Heat Flux Hot Channel Factor, F RTP , K(Z), W(Z), APL NDa "no W(Z)gt for Specification 3/4.2.2, and 0 )
g ",
\
- 6. Nuclear Enthalpy Rise Hot Channel Factor,h' FRTP** d Power Factor Multiplier, MF limits for Specification 3/4.2.3.
g , an } The analytical methods used ta determine the core operating limits shall be j those previously reviewed and approved by NRC in: '
- 1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," i July 1985 (W Proprietary).
(Methodology for Specifications 3.1.1.3 - Moderator Temperature [ N Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux }. Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) (
)
- 2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ -
SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 (W Proprietary). ( (Methodology for Specifications 3.2.1 - Axial Flux Dif ference (Relaxed Axial Offset Contro',) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FgMethodology.)
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", Marcn 1987, (W Proprietary). [
N Luet Aw(Methodology oe w ,4, 1 for Specification 3.2.2 - Heat Flux Hot Channel Factor).
)
/
The core operating limits shall be oeterminea so that all applicable limits (e.g. , fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident ( y analysis limits) of the safety analysis are met. N The CORE OPERATING LIMITS REPORT, including any mid cycle revisions or y supplements thereto, shall be provided upon issuance, for each reload cycle, \ to the NRC Document Control Desk with copies to the Regional Administrator he,-tand ResidentInspector. AR*ew= t 2. SPECIAL REPORTS 6.9.2 Special reports snall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report. McGUIRE - UNITS 1 and 2 6-21 hem :nt. h.'05 (Un" !) l hen &t W. 27 (hit 2-)
for Specification 6'.9.1.9 Attachment 1:
- 4. BAW-10168P,-Rev. 1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," September, 1989 (B&W Proprietary).
.(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)-
- 5. DPC-NE-20llP, " Duke Power Company Nuclear Design Methodology for Core Operating Limits'of Westinghouse Reactors," March, 1990 (DPC Proprietary). l 1
(Methodology for Specification 3.1.3.5 -- Shutdown Rod Insertion l Limits, ;3.1.3.6 Control Bank Insertion Limits, 3.2.1 - Axial 1 Flux Difference, 3.2,2 - Heat Flux Hot Channel Factor, and 3.2.3 ' Nuclear-Enthalpy Rise Hot Channel Factor.)
- 6. DPC-NE-300lP, " Multidimensional Reactor Transients and Safety
-Analysis-Physics: Parameter Methodology," March,-1991 (DPC Proprietary).
(Methodology for Specification- 3.1.1.3 - Moderator Temperature
-Coefficient ,-3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 -
Control 1 Bank Insertion Limits, 3.2.1 --Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel' Factor, and 3.2.3 - Nuclear
-Enthalpy. Rise Hot' Channel Factor.)
- 7. DPC-NF-2010P,." Duke Power-Company McGuire Nuclear Gtation Catawba' Nuclear. Station Nuclear _ Physics Methodology for Reload Design,"
April, 1984. (DPC Proprietary) . (Methodology-for-Specification 3.1.1.3 - Moderator Temperature Coefficient.)- h ihs *E,.- - 5e .5.E.
for Specification 6.9.1,9 :
- Reference 5 is not applicable to target band and APL'd.
- References 4 and 5 are not applicable to W(Z), AP L'3, and W (Z) n,
- Reference 1 is not
- Reference 5 is not applicable applicable to toEbl,'.
- Reference 1 is not
- References 4 and 5 are not applicable to W(Z), AP L'3, and W (Z) n,
FI and MF,,. 6
l l l 2.: T m!MITS // Ij EA5[; 2.1.1 R ArmR CORE The restrictions of this Safety Limit prevent overheating of the fuel n' 'd possible cladding perforation which would result in the release of fission '" products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime' where the heat transfer coefficient is large and the cladding surface
- temparature is slightly above'the coolant saturation temperature. .
Operation above the upper boundary of the nucleate boiling regime couldr,, result in excessive cladding temperatures because of the onset of departurmur from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer-coefficient. DNB is not a directly measurable parameter during operation and m therefore THERMAL POWER and reactor coolant temperature and pressure have been ( 3 related to DNB. This relation has been developed to predict the DNB flux and- ( the locatien of-ONB for axially uniform and nonuniform heat flux distributifins. ) The local DNB heat flux ratio (DNBR), defined as the ratio of the heat flux _ ( that would cause DNB at a particular core location to the local heat flux,Jsc indicative of the margin to DNB. (
)
The ONB design basis is as follows: there must be at least a 95% proba-bility tnat the minimum DNBR of the limiting red during Condition I and II events is creater than or equal to the ONBR limit of the ONB correlation being used (the M correlation in this application). The correlation DNBR set such "* " re is a 95% probability with 95% confidence that DNB will not occur wnen .he minimum DNBR is at the ONBR limit, SWCrny ond the.C4F cocrdobon In meetSg this design basis, uncertainties in plant operat ing parameters, nuclear and thermal parameters, +*e fuel f abrication parametersdare considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. R: e.-. W: in the t e p unt par:m ters re c:ed to ceter " t"e p hnt W unccr ting. ThiONBRuncertainty,-cesineawiv.. ..m arr:1; tion CNEP ' % is usedio establisha&8esi n DNBR value which must be met in plant safety anaiyses using values of input parameters without uncertainties. The c .rces of Figure 2.1-1 show the loci cf ;oints of THERMAL POWER, Reattor
' ;olant System pressure, anc average temperature celow wnich tne calculated DNBR is no less tr.an the cesign DNBR value or the average enthalpy at the vessel exit is ler: than the enthalpy of saturated liquid.
N '5 g The curzes are basec :n a nuclear enthaley rise hot cnannel fattor, F3, of ( M'anc a reference cosinetith a peak of 1.55.te3-ial power sh M An allow- # ( i ce is inci_:ed for an ircrease in F.1"H at red e -- " *ased on the expression: )
,, t.so O/RRu) (
F'
= A 1+X G-m (
)
c'ere s the f ract : d
- f 0 ATED SE;"AL N , and RRH is, opven % h cow. I "c0"::_ - ':~51 M BA2-1 Mena..cm. L. ,: s ii t ---l-)
AM 1 W o c ent N;. 3 (Unit 3
l IU LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued) The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. fos \ T\5 PowerRange,NeutronFlux,HihhRate/ A The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. The Power h e Negative Rat Xerip provide rotection f'o control rod dr% accidents. At igh power, a rod op acciden of a single o multiple rods could cause loca lux peaking whic could cause n unconserva tye local DNBR to N ist. trip will event this f' m o curring trippingthe{thactor.The No Poweh credit ange takenNegative Ra of e for eration Pow Range gqtiveRatetripXorthosecontrol od drop acenignts for wh h DNBR'- willbegKaterthanthedssignlimitDNBRv ue. , Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, rieutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subtritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux chgnnels. The Source Range channels will initiate a Reactor trip at about 10 5 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active, i l l McGUIRE - UNITS 1 and 2 BA2-4 Amendment No.A (Unit 1) Amendment No.- (Unit 2)
bedb - hb% L N! TING SAFETY SYSTEM SETTINGS
\ z 8ASEShithRTOBypassSystem' installed) i
\
Overtemoerature M
\
The Overtemp4rature Delta T trip provides, core protection to prevent DNB: for all comoinationg of pressure, power, coo}4nt temperature, and axial power-distribution, providhtf that the transient slow with respect to piping transit delays from titg core to the tem ture detectors (about 4 seconds), and pressure is within the range betw the Pressurizer High and Low Pressure
- trips. The Setpoint is omatically/ varied with: (1) coolant temperature to:
correct for temperature in
'and includes dynamic compens,ced Atiorytorchanges in density piping delays from and theheat corecapacity of water-to the loop, temperature detectors, (2) presstarizer pressure, and (3) axial power distribu:-
tion. With normal axial pcNer^ddstribution, this Reactor trip limit is always-below the core Safety Limit /s sh' awn in Figure 2.1-1. If axial peaks are greater than design, as indicated dys the difference between top and bottom power range nuclear dete,cfors, the R& actor trip is automatically reduced accoroing to the not ans in Table 2. Overoower AT / The Overpow/ N er Delta T trip provides assuranty! of fuel integrity (e.g. , no-fuel pellet melting and less than 1*. cladding stra(n) under all possible overpower c ditions, limits the required range for Nvertemperature delta i protectio and provides a backup to the High NeutronN1ux trip. The Setpoint is automatically varied with: (1) coolant temperature k correct for tempera-ture ipduced changes in density and heat capacity of wate (2) rate of cnange I of terfperature for dynamic compensation for piping delays f the core to the l Icop' temperature detectors, and (3) axial power distribution, o ensure that I t% allowaole heat generation rate (kW/f t) is not exceeded. Th'ex 0verpower AT t'rio provides protection to mitigate t_he consecuences of various 5 v Secondary
/ breaks Steam Break."as reported in WCAP 9226, " Reactor Core R I
McGUIRE - UNITS 1 and 2 9 2-aa "e nd ent 5.-- % WM+-H
l L!u! TING SAFETY SYSTEM SETTINGS BASES Nith Byoass Svstem Removed: RTOs in Thermowells) Overtemocrature ai The Overtemperature Delta T trio provides core protection to crevent DNB for all comoinations of pressure, power, coolant temperature, ano axial power distribution, provided that the trantlent is slow with respect to Enttes:L rtsponse? bg. delays associated with the RTDs mounted in thermowells, h h * ' :: ::b anc oressure is within the range between the Pressurizer High ano Low Pressure trips. The Setpoint is autcmatically varied with: (1) coolant temoerature to correct for temperature inducea changes in density and heat capacity of water and incluces dynamic compensation for piping delays from the core to the loop temoerature detectors, (2) pressurizer pressure, and (3) axial power distritu-tion. With normal axial power oistribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peans are greater than design, as indicatea by the dif ference between top and bottom-power range nuclear cetectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1. Overoower aT The Overpower Delta T trio provides assurance of fuel integrity (e.g. , no fuel pellet melting and less than 1*. cladding strain) under Fn::::h overoower concitions, limits the required range for overtemperature delta T protection, and provides a bacKuo to the High Neutron Flux trip. 'he Setpoint is automatically varied with: (1) coolant temperature to correct for tempera-ture inouced changes in density and neat capacity of water, (2) rate of change of temoerature for dynamic cc 1censation for instrumentation delays associated ?
*ith the loop temperature detectors, and (3) axial power distribution, to ensure k tnat the allowaole heat generation rate (kW/f t) is not exceeded. The Overpower ATtrip provides protection to mitigate the consecuences of various size steam breans as reported in WCAP 9225, "Deactor Core Response to Excessive Secondary Steam Breat."
l McGUIRE - UNITS 1 and 2 B 2-5 __ _ _ _ ___-Ame
_ . _ _ _ _ _ _ __ . - _ _ _ _ __ _ ~ - _ . _ _ . _ _ _ _ _ _ 1/4. 2 POWER OISTRIBUTION LIMITS Yk b BASE 5 duringThe so.rc1fications Conoi: of this section provide assurancen of fuel i t
. events oy: egrity (1) maintaining the calculated DNBR inrecuency) the core at or1 design limit curing normal operation and in short-term transient s above the the fission erties to within gas release, assumed fuelcriteria.
design pellet v.emoerature, andec cladding anical crop-m ,hano (2) l power constty during Condition I events provides assurance that the i conaitions i w i af 2000"I assumea < - t exceeced. for the LOCA analyses are met nand tial the r eria ECCS a these specifications are as follows:The definit ons of certain hot' channe\ . g g f ~ Heat Flux Hot Channel Factor, is defined as theme
- fQ(qyg) heat flux on the surface of a fuel roc at coreM;;4mese local
-;iwn - CivideQ i
v by the average fuel ances on fuel pellets and roas; rod neat flux, allowing for manufacturing tol er-- l 4 , Nuclear the integral Enthalpy of linearRise powerHot Channel along in: - Factor, is defined asof-the ratio
~
Q6MhY) tewer to the average red power. **e -igr.;;; 6;,,. n;d
/4 2.1 AXIAL FLUX OIFFERENCE k u:t d u!@ B ure beh yC F:(,,y,t) ed %(6y)
The limits on AXIAL FLU be=d rdca ;f it FRTp ' FERENCE (AFD) M that F (Z) REPORT (COLR) ti;.. =.7m 'd 7
, specifiedintheCOREOPgTINGLIMITS <
/ iet ri:' ;; W g #-+ ~ ';jpot exceecea auring either normal power cnanges. operation or 'n the event of xenon reatstributio s m meu ..uu.c usuo. % <.p pew elete sf eu t,cd .a W uut hu bm adped iarget flux of fference is determined at eauilibriu m xenon concittons.
their respective insertion limits and should be inserted n I position for steaoy state operation at high power levels. normal target flux difference obtaineo under these concitions The divided value of theby th , of RATED THERMAL POWER is the target flux of fference e fraction at RATED THERM for tne _ associated core burnuo concitions. Target flux differences f THERMAL F0WER 'evels by the approcriate fractional THERMAL POWER level. are cotainea by multiolying tne RATED or otner value THERMAL P 3 the target flux difference value is necessary to reflect core burnuoThe periodic updat nsiderations.
\
k pa ) a ca ee,
@ o 4 -mJ ud rg (x,yg) I; ;t for <
pcs hu3h4. McGUIRE - UNITS 1 -ene-e- 8 3/4A2-1 7.;r mi ue '00 ' Unit 1)
.- n-.. n o ... 67(Uri )
l l UA N b POWER DISTRIBUTION LIMITS BASE 3 XIAL FLUX DIFFERENCE (Continueo) CI'
^ ,
At power levels celow APL'NO _\ i.e. that cefined oy the RAOC operating proceaure and limits.. the limits on AFOl\are defined i These limits ere calculated in a manner sucn that expected operational transients, e.g. loao follow limits. operations, would not result in the AFD deviating outside of those ' However, in the event sucn a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in I significant xenou redistribution such that the envelope of peaking factor would power level. change suf ficiently to prevent operation in the vicinity of the APL'$D At. power levels greater than APLND ,
- 1) RAOC, the AFD limits of whicn are definea in the COLR, and 2) base loadtwo moces of co l coeration, wnien is cefinea as the maintenance of the AFD within a COLR
' specified band aoout a target value. The RAOC operating procedure above APLNO is the same as that defined for operation below APLNO However, it is possible when following extenced loaa following maneuvers that the AFO limits may result in restrictions in the maximum allowed power or AFD in oroer to guarantee operation with Fg (:) less than its limiting value. To allow operation at the maximum .'
l permissible value, the case load operating procedure restricts the indicateo AFD to relatively small target banc ano power swings (AFD target band as specifiec { in the COLR, APL ND 3 j f' For base load operation, it is exoected that the plant will !operate ~ I target bano. Operation ous1de of the target band for the snort time perica
' allowed will not result in significant xenon recistribution sucn that the enve)oce of peaking f actors would cnange suf ficiently to pronibit continueo operation in the cower region definea above. , To assure there is no residual xenon recistri-Dution impact from cast coeration on the base load operation, a 24 hour waiting i cerica at a power level above APL"O and alloweo by RAOC is necessary. During tnis time ceriod load changes
( loao procecure. After the waiting period extenced and rod motion are restrictec to that allowed by the case x base load operation is permissiole.
-s The comouter catemines the one minute average of eacn of the OPERABLE excore catector outouts and provices an alarm message imeoiately if the AFD , ,,
for at i WL J least 2 of 4 or 2 of 3 OPERABLE excore enannels arel DJbutsice the
,sTiowea al power operating spaceh(for RA^C swciau m.,, o L vuuida se
, ah:: ^7 + = = + W ( for ;eae i s.J uw..
R e, ). These alarms are active 1.nen power is greater than,( +) 50% of RATED THERMAL POWERc(for RAOC vw=r au smr r T "LND , , r . c mow ,. ,, , , o i. a n ; . 'cw ty ce tien g e m m, m 7, 7. ..m ;eae taac
.m r m nistre ::;g m, re iti0r " ' ' f 4 '* a ^f *ha 6. . m mu rt ;;erToa ci time curing woiw i
t irg;'. 0;.iw na allw ;e
\
cluriq nor m.g po,ar oper6m i S. ; pfl 1 ucGUIRE - UNITh 1 -afe-e 3 3/4A?-2 ' : x .= ": d"nM U M:ac;r.eu s Hv. 3 En i i --2 ) _ _ _ _ _ _ . _ . _ . _ _ . _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - ~ ^ ^ -
L% T 1 POWER DISTRIBUTION LIMITS BASE! 3/4.2.2 ana 3/4;2.3 HEAT FLUX HOT CHANNEL FACTOR NUCLEAR ENTHALPY RISE n07 CHANNEL F ACTOR sad $"E "lU "TE AND ( W.4 ) The limits on heat flux hot channel f actor, " -v "+= enthalpy rise hot enannel factor ensure that: and nuclear (1) the cesign limits on peak pC a LOCA i etu : local thepower ee 'c:' t **m and minimus density i ONBR are not exceeded, and (2) i
- t c:r d th 22^^*P ECCS acceo-tance REPORT criterla}Mett. 4eeeglimits are specified in the CORE OPERATING LIMITS fication 6.9.1. (s g,Q b
as specified in Specifications 4.2.2 and 4.2.3.A -Ech n n:n i; measurable sufficient to insure that the limits are maintained orovir'a+This periodic surveillance is
$ w A h + cAuno 2 A.ne
.antrol roos in aem n aam unwy rise he ewnel Acht are <* c0 single group move to insertion ciffering by more than @ gether steos fromwiththe no group inoivicual-demano rod position; g g og b.
Control red groups in Specification are sequenced with overlapping groups as described 3.1.3.6; c. The control rod insertion limits of Specifications 3.L3.5 and 3.1.3.6 are maintainec; ano d. The axial power distribution, expressed in terms of AXIAL FLUX OIFFERENCE, is maintained within the limits. res (67) q h > w Fy ill be maintained within its limits orovided Conditions a. through above _ e maintainsa.J As noted on the figure specifiec in the CORE OPERATING (LIMITS REPORT (COLR), RCS flow rate and power may be " traced off" against o
$ another (i.e. , a low measureo RCS flow rata is acceptable if the power level is ldu decreased) to ensure that the calculatea ONBR will not be celow the design DN8R }
W 6value. The relaxation of F'y M as a function of THERMAL POWER allows changes in Qhe racial power shape for all permissible rod insertion limits. gfi"* ";- R as calculatea in Specification 3.2.3 and used in the figure specified - in the COLR, accounts for Fh less than or equal to the F[TP limit specified in [ j theCOLR.ThisvalueisusedinthevariousaccidentanalyseswnereFh
! influences parameters other than DNBR, e.g. , peak clad temoerature, and thus is 1 the maximum "as measureo" value allowed.
Margin between the safety analysis limit ONBRs and the design limit DNBRs is maintained.
' l tion core DN8R penalty (2%) and the appropriate fuel rod/ bow ON ; (WCAP - S691, Rev. 1). '
w[- When an F j , q t measurement is taken, an allowance for totn exoerimental error jqi , m.w ' ( and manufacturing tolerance must be made. C An allowance of 5% is appropriate
.5 McGUIRE - UNIT $ 1 ne e/we+ <, 8 3/4f,2-2a
".u ~ .. nu.105Gy L 1)
' \Minassas 6 av. O / WG 3 4- 'i
l for Power Distribution Limits Bases Attachment 1: The limits on the nuclear enthalpy rise hot channel fcctor, Fy(X,Y), are specified in the COLR as Maximum Allowable Radial Peaking (MARP) limits, obtained by dividing the Maximum Allowable Total Peaking (MAP) limit by the axial peak (AXIAL (X,Y)) for location (X,Y) By definition, the Maximum Allowable Radial Peaking limits will result in a DNBR for the limiting transient that is equivalent to the DNBR calculated with a design Fy(X, Y) value of 1.50 and a limiting reference axial power shape. For transition cores, MARP limits may be applied to both Mark-BW and optimized fuel types provided allowances for differences in DNBR are accounted for in the generation of MARP limits. The MARP limits specified in the COLR include allowances for mixed core DNBR effects. The relaxation of Fy(X,Y) as a function of THERMAL POWER allows for a change in the radial power shape for all permissible control bank insertion limits. This relaxation is implemented by the application of the following factors: k= [1 + (1/RRH) (1 - P)] where k - power factor multiplier applied to the MAP limits P - THERMAL POWER / RATED THERMAL POWER RRH is given in the COLR A
l l POWER DISTRIBUTICH LIMITS VN IT .1 v BASES HEAT M0T FLUXFWOT CHANNEL ACTORCHANNEL ( Cnnn n"-a ' FACTOR ana RCS FLOW RATE AND NUCLEAR ENTHALPY RISE
< for a full-core mao taken with the Incore Octector Flux Mapping System, anc L
- a 3% allowance is accropriate for manufacturing tolerance. j N
When RCS flow rate and F are measured, no additional allowances are 7*'a' f*NJ ** s.m 6, necessary prior to comoarison with the limits of the figure specifiec in the 1.s t " COLR. Measuresent errors of 1.7% for RCS total flow rate and 4% for FN have been allowea for in detemination of the design DNBR value. # { The measurement error for RCS total flow rate is based upoa pr.rfoming a precision neat calance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the femowater venturi wnich might not be detected could bias the result from the precision heat balance in a non- i conservative manner. Therefore, a penalty of 0.1% for uncetected fouling of the fencwatar venturi is included in the figure specified in the COLR. Any i l fouling which might bias the RCS flow rate measurement greater snin 0.1% can be detectea by monitoring and trencing various plant performance paramstars. !
\ If detected, action shall be taken esfore performing suosequent precision h(at (
balance measurements, i.e. , either the effect of the fouling shall be quantifiec ! and comoensated cleanea for the to eliminate in the RCS flow rate measureevnt or the venturi shall be fouling. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degracation which could lead to operation outside the accent - able recion of coention specified on the figure soecified in the mo l [ The not enannel factor (z) is measured periccically and increased by
/ cycle and heignt caoencent power f actor appropriate to either RAOC or base
\
loao operation, W(z) or W(z)gg, to provide assurance that the limit on the hot channel f actor, F q (z), is met. W(z) accrunts for the effects of nomal ( operation transients and was determinea from expected power control maneuvers over the full range of burnup conaitions in the core. W(z)gt accounts for the more restrictive operating limits allowea by case leaa operation wnicn result in less severe transient values. The W(z) function for normal operation and the W(z)gg function for base loac operation are specified in the CORE OPERATING LIMITS REPORT per Soncification 6,9.1.3. lue f.Mn< .A 2- ) J
,' W2 WeGUIRE - '.mIT! 1 --- 2 5 3/4/2-4 ^-mt t. OGCi t U -
l MOM
i for Power Distribution Limits Bases Attachment 2: The hot channel factor, F@ ( X , Y , Z ) , and the nuclear enthalpy rise hot channel factor, FL (X, Y) , are measured periodically to verify that the core is operating as designed. F8(X,Y,Z) and FL (X, Y) are compared to allowable limits to provide reasonable assurance that limiting criteria will not be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable Control Assemblies) , 3.2.1 (Axial Flux Difference), and 3.2.4 (Quadrant Power Tilt Ratio). A peaking margin calculation it performed to provide the basis for decreasing the width of the AFD and f(AI) limits and for reducing TilERMAL POWER. When an F;(X, Y, Z) measurement is obtained from a full-core map in accordance with surveillance requirements of Specification 4.2.2, no uncertainties are applied to the measured peak since a measurement uncertainty of 5.0% and a manufacturing tolerance of 3.0% are included in the peaking limit. When F[(X, Y, Z) is measured for reasons other than meeting the requirements of Specification 4.2.2, the measured peak is increased by the radial-local peaking factor and appropriate allowances for measurement uncertainty and for manufacturing tolerances. When an FL(X, Y) measurement is obtained from a full-core map, regardless of the reason, no uncertainties are applied to the measured peak since the required uncertainties are included in the peaking limit.
\ l Y$IT I 80VER Of5TRIBUTION LIMITS BASE 5 3/4 L 4 OUADRCe, POWER TIL* 2ATiO The QUADRANT POWER TILT RATIO limit assures that ne racial cower 01str1-tution satisfies the oesign values usea in tne cowsr capanility analysis. Racial power c1stribution measurements are mace curing STARTUP testing anc periccically during power operation. C t" " T (DY tnan The 2-nour time allowance for operation with a tilt concitten greater 1.02 but less than 1.09 is provideo to allow identification ano correc-tion of a croppea or misaligneo rec. In the nt sucn action oces not cor-rect the tilt, the margin for uncertainty on reinstatea by reaucing h( the wer ey 3% from RATED THEFJtAL POWER for eacn percent of tilt in excess Q
?f . 0.
I ,2.0 % For purcosas of monitoring 00ADRANT POWER TILT RATIO w- s etector is inoperable, the moveaole incere aetectors ar* .a that tne normalizac symetric ;cwor cistribution is consistent, ., AN T POWER TILT RATIO. The incere detector monitoring is cone , , f'n i imer* , flux man or two sets of four symmetric tnimalesy The two sets of four synenetric A.n1 motes is a unique set or eignt catector tocat1ons. These locations are
-8 E-5 E-ll, H-3, n-13, L-5 L-ll, N-6.{ q j
3/4.2.5 DN8 PARAMETERS Dew " The I'mits on the DNB-relateo parameters assure that eacn of the cara-meters are maintained witnin the normal steacy state enveloce of operation f y assumeo in the transient ano accioent analyses. with tne initial FSAR assumotions and have oeen The ilmits areaesonstrated analytically consistent (Nu 2.) aceouatetomaintainacasignlimitONBRthrougneuteacnanalyzeotransient.[\ The inoicateo T gg values ano tne inoicated oressurizer pressure values
- orresoona to analytical liraits of 592.6'F ana 2220 psia resoectively, itn f.TawA 1, e allowance for inoication instrumentation measurement uncertainty. A
-^ .
AF=La ! The 12-hour periccic surveillance of these carameters througn instrument reacout is sufficiont to ensure that the carameters are restorea witnin tnetr limits following loao enanges ano etner expectea transient operat1on. Incication instrumentation measurement -uncertaintles are accounted for in tne limits orovicea in Table 3.2-1. wc GUIRE - UNITS 1 ano 2 B 3/4A2-5 4memseen.t "c.6', (U'"t
-i
'"M* "
Anet Nud
for QPTR bases : The limic of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with the x-y plane power tilts. The peaking increase that corresponds to a QUADRANT POWER TILT RATIO of 1.02 is included in the generation of the AFD limits. 4
. __ _ ._ _ _ - _ _ _. _ ._,._m _ _ _ . _ _ _ . ._.
1 l for 3/4.2.5 BASES
-Attachment 2:
As;noted or. Figure 3.2-1, RCS flow rate and THERMAL POWER may be
" traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the power level is decreased) to ensure that the
. calculated DNBR will not be below the design DNBR-value. The relationship defined orf Figure' 3.2-1 remains valid as long as the 4 limits placed on_the nuclear enthalpy rise hot channel factor, l Fu (X, Y) , in Specification 3.2.3 are maintained. !
l
- u i
i Attachment 3: When RCS flow rate is measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-1 since a measurement error of.l.7% for RCS total flow rate has been allowed for in determination of the design DNBR value.
'The measurement e'ror-for r RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS i flow rate indicators. Potential fouling of the feedwater venturi which might net be detected could bias the result-from the precision
. heat balance 2n a non-conservative manner. Therefore, a penalty of 0.1% for undetected-fouling of the feedwater venturi is included in Figure _3.2-l'. -Any; fouling which might bias the RCS flow rate measurement greater than 0.1% can.be detected by monitoring and trending.various plant performance parameters. If detected, action shall-be-taken before performing subsequent precision _ heat balance measurements, i.e., either-the effect of the fouling shall be
_ quantified and compensated for in the-RCS flow rate measurement or the
. venturi shall be cleaned to eliminate lthe' fouling.
~. _ .,. . __
IT} 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design limit during all normal operations and antici-pated transients. In MODES 1 and 2 with one reactor coolant loop not in oper-ation this specification requires that the plant be in at least HOT STANDBY within 1 hour. In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat; however, single failure considerations the u n c.od r o h dtkun k require m %d.c w that three :.loops
-d-9com e.v-ebe9 om OPERABLE.
cc o v- so Also,ber&qc4 a.psuges do c-In MODE 4, and in MODE 5 with reactor coo 18$t foops fiMed, a hngle reactor coolant loop or RHR 1000 provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat remnval capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. .The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting a reactor coolant pump with one or more RCS cold legs less than or equal to 300 F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by . .her: (1) restricting the water volume in the pressurizer and thereby providing a volume for the reactor coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water tempera-ture of each steam generator is less than 50 F above each of the RCS cold leg temperatures. Amenament Mc M Unit U McGUIRE - UNITS 1 and 2 8 3/4A4-1 AmenaEnt No M unit C
LatnT A 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T shall not exceed the limits shown in Figures &:-lM and 2.1-2 for four and We)e loop operation, respectively.
" L l-l b APPLICABILITY: MODES I and 2. (()Al)T L (pp/L/ )
ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour, and comply with the require-ments of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4, and 5 ( UNTS /ne'36-Q ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour, and comply with the requirements of Specification 6.7.1. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1. McGUIRE - UNITS I and 2 2h
i l UN /T z ! r. 678 FLOW PER LOOP = 96250 gpm
~
UNACCEPTABLE OPERATION 2400 sg . Ds te n M- E000 A y S 1a
~
us -
? lI?S
* # #/
e- gig . 4 600 -
- ACCEPTABLE OPERATION SM -
388 e i i 6 e i i i e a i i 8 3 # le W IN -120 POWER (PERCENTAGE OF NOMINAL) e FIGURE 2.1-lh UNITO O h
- 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION McGUIRE - UNITS 1 and 2 2@a Amendment No.ll(Unit 1)
Amendment No.98(Unit 2)
s bl[ 1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS .
2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-lh.[ g y g g APPLICABILITY: AsshownforeachchannelinTable3.3-lh[hOy ACTION: With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value. McGUIRE - UNITS 1 and 2 2h
~
kW .; e , t 3 TABLE 2.2-Ib ' a c- REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i =
- 7. FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E 1. Manual Reactor Trip N.A. N.A,
- .-e E
[ 2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED. Low Setpoint - < 26% of RATED
., . THERMAL POWER THERMAL POWER High Setpoint High Setpoint - < 110% of RATED 1 109% of RATED THERMAL POWER THERMAL POWER 4
t
- 3. Power Range, Neutron Flux, 1 5% of RATED THERMAL POWER with $ S.5% of RATED THERMAL POWER
, High Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds ;
- 4. Power Range, Neut'on t Flux, 1 5% of RATED THERMAL POWER .with < 5.5% of CATED THERMAL POWER High Negative Rate a time constant 1 2 seconds with a time constant 2 2 seconds l
h 5. Intermediate Range, Neutron Flux i 25% of RATED THERMAL POWER 1 30% of RATED THERMAL POWER fr
- 6. Source Range, Neutron Flux i 10 5counts per second i 1.3 x 105 counts per second
- 7. Overtemperature AT See Note 1 See Note 3 $
- 8. Overpower AT- See Note 2 See Note 4 D j
l (( 9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig
- 10. Pressurizer Pressure--High 1 2385 psig i 2395 psig OO t
~~
. g 11. Pressurizer Water Level--High -< 92% of instrument span -< 93% of instrument span -
j 2
.e os > 88.8% of minimum measured f I 12. Low Reactor Coolant Flow > 90% of minimum measured 22 Tiow per loop
- flow per loop *
- i. rr11
! y
- Minimum measured flow is 96,250 gpa per locp. q=>
. I l
(Urxrz TABLE 2.2-1MContinued) Y [f glgg j I, c REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 A - TRIP SETPOINT ALLOWABLE VALUES
, FUNCTIONAL UNIT 3 12% of span from 0 to 30% of 3 11% of span from 0 to 30% of k 13. Steam Generator Water RATED THERMAL POWER, increasing RATED THERMAL POWER, increasing g Level--Low-Low linearly to 3 40% of span at to 39.0% of span at 100% of g RATED THERMAL POWER.
100% of RATED THERMAL POWER
- 14. Undervoltage-Re:ctor 1 5082 volts-each bus 1 5016 veits each bus
[ Coolant Pumps
- 15. Underfrequency-Reactor 3 56.4 Hz - each bus 3 55.9 Hz - each bus Coolant Pumps l 16. Turbine Trip Low Trip System Pressure 3 45 psig 1 42 psig g<n a.
- b. Turbine Stop Valve Closure 1 1% open 1 1% cpen N.A. N.A.
- 17. Safety Injection Input from ESF
- 18. Reactor Trip System Interiocks Intermediate Range Neutron Flux, P-6, 1 1 x 10 20 amps 1 6 x 10 xx amps gg a.
Enable Block Source Range Reactor Trip gg EE Low Power Reactor Trips Block, P-7 gg b. c> e 10% of RATED ? 9%, < 11% of RATED yg 1) P-10 Input THERMAL POWER THERMAL POWER
/ ( 10% RTP Turbine < 11% RTP Turbine cc 2) P-13 Input Impulse Pressure Impulse Pressure Equivalent Equivalent UU
NW {
- TABLE 2.2-IkContinued)
@ REACTOR TRIP SYSTEM INSTRUMENIATION TRIP SETPOINTS rn FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E
; c. Power Range Neutron Flux, P-8, < 48% of RATED
< 49% of RATED
~
Low Reactor Coolant Loop Flow, THERMAL POWER THERMAL POWER and Reactor Coolant Pumn Breaker
$ Position cs
- d. Low Setpoint Power Range Neutron 10% of RATED > 9%, i 1E% of RATED Flux, P-10. Enable Block of THERMAL POWER THERMAL POWER Source Intermediate and Power Range Reactor Trips
- e. Turbine Impulse Chamber Pressure, j
P-13. Input to Low Power Reactor 5 10% RTP Turbine $ 10% RTP Turbine l Trips Block P-7 Impulse Pressure Impulse Pre:sure ;
}* Equivale<' Equivalent
- 19. Reactor Trip Breakers N.A. N.A.
1
- 20. Automatic Trip and Interlock logic N.A. N.A.
l l I t [ t
U uri~ 2 t c IA81E 2.2-Ab(Continued) i f I REACTOR 1 RIP SYST[M INSTRuffMTATION TRI.* SETPOINIS r
=
NOIATION E
] NOTE 1: OVERTEMPERATURE AT 1+t 5 1
) (g ,1 T2 S) i 33o gg l _g I l+t S IEII I+1 '
AT(f*, 2 6 53 'I 3 * "3(P-P') - f g(al)l n GOP!]ASO,) kpr6G y Where: AT = Measured AT by Ris = n.feTo 1..n . -
-*iaa. '
4 1 + t'5 = Lead-lag compensator on ocasured AT, 3 ,g 13, 12 = Time constants utilized in the lead-lag controller for l j al , 1 > 8 sec., r2 5 3 S*C-. I r n ; 3
@ 3,
= Lag compensator on measured AT, #
i ! ! 13 = Time constants utilized in the lag compensator for ai, 13$ 2 sec. l f . AT = Indicated AT at RATED THERMAL POWER, ! ! L K g
< 1.200, ;
'l(( : ,33 K = 0.0222 2 33 k 1 + t,5
" =
3,,g The function generated by the lerd-lag controller for T,,g dynamic compensation, .b to* is = Time constants utilized in the lead-lag controller for T
^"9,
- t. > 28 sec, is $ 4 sec.,
cc' jss .
,+ e+
I = Average temperature. *f, l r Og 7,
=
Lag compensator on measured I, 3 , i I. ' l i; I i
mi 2_ TABLE 2.2-d(Continued) E REACTOR TRIP SYSTEM INSTRLHENTATION TRIP SETPOINTS Si N CTATION (Continued) C 5 NOTE 1: (Continued) d = Time constant utilized in the measured T,yg lag compensator, is i 2 sac g Ts at RATED THERMAL POWER, k T' =
$ 588.2'F Reference T,yg
" = 0.00 W 5, K
3 P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure), S = Laplace transform operator, sec 2, and f3(al) is a function of the indicated difference between top and bottom detectors lh of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for q t ~4 b between -29% and +7.0%; f (al) = 0, where qt and a are percent RATED 1 b l THERMAL POWER in the top and bottom halves of the core respectively, and q t *U b EF is total THERMAL POWER in percent of RATED THERMAL POWER; gg 55 g exceeds -29%, the AT Trip Setpoint .l gg (ii) for each percent that the magnitude of qt b
"" shall be automatically reduced by 3.151% of its value at RATED THERMAL POWER; and &F q exceeds +7.0%, the AT Trip Setpoint l
(, (iii) for each percent that the magnitude of qt b
"*M shall be automatically reduced by 1.50% of its value at RATED THERMAL POWER.
22
=_' E.
t+ c*
F. l
.l i: M b . .
<f ' TABLE'2.2-lb(Continued) - ' i !' j!f REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS t 4 m .
-5 NOTATION (Continued) i
'In .
I
.. NOTE 2: OVERPOWER AT :
c > i- g 1. ,'3 1 1,5 1- 1 i Z,
-K v A T (1 + T 25) (1 + TsS) < ATo-{K4 5 (1 + 1 7 5) (1 + r.5) T -K6 [T(1 + r.S)- T"] - f2 (al)]
w
- .. Where:- :AT = As defined in Note 1,
=
a 1+r5 ! y y = As defined in Note 1 ! ri, 12 - = As defined in Note 1 I 1+153 = A,s defined in Note 1, i
= As defined in Note 1, I m AT, i: Do ~>
l K < 1.0900, 4 (, Kg = 0.02M for knasing average twerature and 0 for & creasing averay . temperature. I
- l. '
t !. 15 7 ' l
= '
li' 7 . compensation, The function generated by the rate-lag controller for T,yg dynamic j
=
_ , _ , .12 Time constant utilized in the rate-lag controller for T,yg,1 7 > 5 sec, i
@@ 1 g g. y.
= As defined in Note 1, ne 63 zz
- [ P . r. = As defined in Note 1,
! .AN !
- j. 22 K 6
=
0.00169/*F for.T > T" and K6 = 0 for T $ T", i i: 11 !
~ ~ . ;
t, i vv . i 1 l.
I % i i x TABLE 2.2-Ib(Continued) ! 3- n- != E. . REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS, + g: NOTATION (Continued) I. . I
-g I i- U' w
T = As defined in Note-1,
- " T" =
-< 588.2*F Reference Tavg at RATED THERPML POWER, l a
.,. t
!- .5- = t.s defined in~ Note 1, and I , m ff(AI) = 0-for all al. Note 3: The channel's' maximum'. Trip Setpoint shall not. exceed its computed Trip Setpoint by.more than l [
, 3.6% of Rated Thermal Power.
s Note 3a: Thc channel's maximum Trip Setpoint shail not exceed its computed Trip setpoint by core than 2%. , 3 t
- 3. l
- . .- Note'4
- The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 4.2% 7 t of Rated Thermal Power, !
l b f
\
,. pp.
!b i- aa i
- i. 22
- i. aa i
' II
[h n 22 i AA j
- ~~ t . I i: f i .
Yr k 3/4.2 ' POWER DISTRIBUTION LIMITS i 3/4,2.1 AXIAL FLUX O!FFERENCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:
- a. the allowed operational space as specified in the CORE OPERATING LIMITS REPORT (COLR) for RAOC operation, or [
e .
- b. the target band specified in the COLR about the target flux difference during base load operation.
-APPLICABILITY: MODE-1 above 50% of RATED THERMAL POWER *. INT d ON ACTION: !
- a. For RAOC operation with the indicated AFD outside of the limits ,
specified in the COLR, k"g=
- 1. Either restore the indicated AFD to within the COLR 3 limits within 15 minutes, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -
High Trip setpcints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
- b. For base load operation above APLND" with the indicated AXIAL FLUX i OIFFERENCE outside of the applicable target band about the target flux difference:
- 1. Either restore the indicated AFD to within the COLR specified target bsnd limits within 15 minutes _or
}
NU
- 2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes,
- c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the -
J;, COLR.
"See Special-Test Exception'3.10.2.
ND
**APL is the minimum allowable (nuclear design) power level for base load e operation and is specified in the CORE OPERATING LIMITS REPORT per p Specification 6.9.1.9.
McGdIRE - UNITS 1 and 2 3/462-1 AmendmentNo.N(Unit 1) Amendment No j B7\(Unit 2)-
06 5 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- 2. At least once per hour for the first 24 hours after restoring the AFD Monitoring Alarm to OPERABLE status.
- b. Monitoring and log 0 ing the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereaf ter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits. 4.2.1.3 When in Base Load operation, the target axial flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 When in Base load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference in conjunction witn the surveillance requirements d of Specification 3/4.2.2 or by linear interpolation between the most recently ( 3 measured value and the calculated value at the end of cycle life. of Specification 4.0.4 are not applicable, The provisions \- McGUIRE - UNITS 1 and 2 3/482-la AmendmentNoh(Unit 1) AmendmentNo/q(Unit 2)
. . - . . - - ._ - -. - . - .. - - . - . . _ _. - - - . ~ . . - . - . . . . . - . - . - - l i i POWER DISTRIBUTION LIMITS ! l 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be limited by the following relationship: RTP F9 (I) 1 F Q K(Z) for P > 0.5 [, P RTP Fq (Z) $ F Q K(Z) for P 5, 0.5 _Y,
- 0. 5 Where FRTP = the F limit at RATED THERMAL POWER (RTP) specified '
- 0 q in the CORE OPERATING LIMITS REPORT (COLR), {;3 THERMAL POWER , and '
P = RATID THERMAL POWER K(Z) = the normalized Fq (Z) for a given core height ; specified in the COLR. APPLICABILITY: MODE 1.(U/)/72.@jk[ l ACTION:. With Fq (Z) exceeding its limit: a._ Reduce THERMAL POWER at least 1% for each 1% Fn(Z) exceeds the limit within 15 minutes and similarly reduce the PowVr Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may preceed for up to a total of 72 hours;. subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K 4) have been reduced at least 1% (in AT span) for each 1% F q (Z) exceeds the limit; and
- b. Identify'and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTTON a.. above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. q McGUIRE - UNITS 1 and 2 3/432-6 Amendment No Unit 1)
Amendment No. (Unit 2)-
. _ x _ -_ _ . _ _ _ _ _ . _ ~ _ . ~ . - - - -- - - - - ~ ~
_ _ _ _ _- - _._ _ _ _ . _ _ _ _ --.___ _ _ _ m.._ ___ __. ._ . Vuri 2 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.? For RAOC operation, F (z) shall be evaluated to determine if F q(z) is within its limit by: 9 {
- a. Using the movable incore detectors to obtain a power distribution I map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. I
- b. Increasing the measured qF (z) component of the power distribution map by 3% to acccunt for manufacturing tolerances and further y.
increasing the value by 5%** to account for measurement uncertainties, i Verify the requirements of Specification 3.2.2 are satisfied. I
- c. Satisfying the following relationship:
RTP F F9 "(z) i Q x K(z) for P > 0.5 P x W(z) RTP N F Fq (z) 1 Q x K(z) for P < 0.5 W(z) x 0.5 - where F (z) is the measured F (z) increased by the allowances f.* , 9 manufacturing tolerances and measurement uncertainty, is the FhTP F limit, K(z) is the normalized F (z) as a function of core height, q 9 P is the relative THERMAL POWER, and W(z) is the cycle deoendent function that accounts for power distribution transients encountered RTP during normal operation. F g
, K(z), and W(z) are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9.
N
- d. Measuring Fq (z) according to the following schedule:
- 1. Upon achieving equilibrium conditions af ter exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or 9
- 2. At least once per 31 Effective full Power Days, whichever occurs first.
*During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
**For Unit 1, Cycle 7, when the number of available moveable detector thimbles g is greater than or equal to 50% and less than 75% of the total, the 5% mea- ,
surement uncertainty shall be increased to [5% + (3-T/14.5)(2%)] where T is p the number of available thimbles. McGUIRE - UNITS 1 and 2 3/402-7 Amendment No) 7(Unit 1) Amendment No Unit 2) __ . ~ _ _ _ _ _ -- _ _ m.. _ _- _ _ . _ . _ . _ _ _ _ . . _ _ _ . _ _ - _ - _ - . _ _ __
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- e. With measurements indicating maximum [FN (z) over z (K(z)j N
has increased since the previous determination of Fq (I) either of the following actions shall be taken: N
- 1) F q (z) shall be increased by 2% over that specified in Specifi-cation 4.2.2.2c. or N
- 2) F g (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that is not increasing.
maximum Ih(z) F over z ( ~K (z) j
- f. With the relationships specified in Specification 4.2.2.2c. above not being satisfied:
- 1) Calculate the percent F q (z) exceeds its limit by the following expression: ,
maximum p H q (z) x W(z)
\-1dx100 for P -
0.5 (overI p p x K(z) - s q%
' 5 I 9
'[ maximum M
)
- WI ) 3
-1 -
100 for P < 0.5 over z R Y {h,5 x K(z) /
)
- 2) One of the following actions shall be taken:
a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits of Specification 3.2.1 by 1% AFD for each percent Fn(z) exceeds h its limits as determined in Specification 4.2.2.2f.1). Within 8 hours, reset the AFD alarm setpoints to thesa modified limits, or b) Comply with the requirements of Specification 3.2.2 for F9 (z) exceeding its limit by the percent calculated above, or c) Verify that the requirements of Specification 4.2.2.3 for base load operation are satisfied and enter base load operation. McGUIRE - UNITS 1 and 2 3/402-8 AmendmentHo1%(Unit 1) Amendment No/37(Unit 2)
POWER DISTRIBUTION L!MITS MhT 2 WRVEILLANCE REQUIREMENTS (Continued)
- g. The limits specified in Specifications 4.2.2.2c, 4.2.2.2e., and 4.2.2.2f.
above are not applicable in the following core plane regions: 1, Lower core region from 0 to 15%, inclusive.
- 2. Upper core region from 85 to 100%, inclusive.
4.2.2.3 Base load operation is permitted at powers above APLN0* if the following conditions are satisfied:
- a. Prior to entering base load operation, maintain THERMAL POWER above ND APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours. Maintain base load operation sur-veillance (AFD within the target band about the target flux difference of Specification 3.2.1) during this time period. Baseloadoperatig is then sg m itted providing THERMAL POWER is maintained between APL and APL ' or between APL NO and 100% (whichever is most limiting) and FQ g rveillance is maintained pursuant to Specification 4.2.2.4.
APL is defined as: RTP Q APL OL =over minimum Z [ (F x K(Z) ) x 100% F (Z) x W(Z)BL where: F (z) is the measured Fq (z) increased by the allowances for manufacturing tolerances and measurement uncertainty. FfTPis the F q limit. K(z) is the normalized Fq (z) as a function of core height. W(z)gg is the cycle dependent function that accounts for limited power distrPution transients encountered during base luad operation. Ff,8(:.),i.ndW(z)BL are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.o.
- b. During base load operation, if the THERMAL POWER is decreased below ND APL then the conditions of 4.2.2.3.a shall be satisfied before re-entering base load operation.
4.2.2.4 During base load operation F q (Z) shall be evaluated to determine if F (Z) is within its limit by: q
- a. Using the movable incore detectors to obtain a power distribution ND map at any THERMAL POWER above APL ,
- b. Increasing the measured Fq (Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5%** to account for measurement uncertainties. %
Verify the requirements- of Specification 3.2.2 are satisfied. ND
*APL is the minimum allowable (nuclear design) power level for base load operation in Specification 3.2.1. '
**For Unit 1, Cycle 7, when the number of available moveable detector thimbles is greater than or equal to 50% and less than 75% of the total, the 5% mea-surement uncertainty shall be increased to (5% + (3-T/14.5)(2%)] where T is Ct the number of available thimbles.
McGUIRE - UNITS 1 and 2 3/4J2-9 Amendment No.1 Unit 1) Amendment No. RUnit2)
POWER DISTRIBUTION *.!MITS ///T ( SURVEILLANCE REQUIREMENTS (Continued)
- c. Satisfying the following relationship:
RTP C - 0 for P > APL ND I s F (Z) 5,p l M where: F(Z)isthemeasuredFg(Z). q F is the F q limit. ( K(Z) is the normalized Fq (Z) as a function of core height. P is the b relative THERMAL POWER. W(Z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during base load operation. F TP , K(Z), and W(Z)BL are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. '
- d. Measuring F (2) in conjunction with target flux difference deter-mination according to the following schedule:
- 1. Prior to entering base load operation af ter satisfying Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been ND maintained above APL for the 24 hours prior to mapping, and.
- 2. At least once per 31 cffective full power days,
- e. With measurements indicating F (Z) maximum [
over Z y] has increased since the previous determination F (Z) either of the following actions shall be taken: N
- 1. F (Z) shall be increased by 2 percent over that'specified in 4.2.2.4.c, or
- 2. F (Z) shall be measured at least once per 7 EFPD until 2 successive maps indicate that maximum ( ) is not increasing, over z l- f. With the relationship specified in 4.2.2.4.c above not being satisfied, either of the following actions shall be taken:
- 1. Place the core in an equilbrium condition where the limit in 4.2.2.2.c is satisfied, and remeasure FqM (Z), or l
~
l McGUIRE - UNITS 1 and 2 3/4ff-9a AmendmentNo}(/XUnit1) l AmendmentNo./gfunit2) - J
1 . b L POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
- 2. Comply with the requirements of Specification 3.2.2 for F (Z) ;
n exceeding its 1.imit by the percent calculated with the following ; expression: , N [(max. over z ofgp[ F (Z) x W(Z)BL ] ) -1 ] x 100 NDfor P > APL , F y n ; p x K(Z)
- g. The limits specified in 4.2.2.4.c, 4.2.2.4.e. and 4.2.2.4.f above
-are not applicable in the following core plan regions:
- 1. Lower core region 0 to 15 percent, inclusive.
- 2. Upper core region 85 to 100 percent, inclusive. !
4.2.2.5 When Fg(Z) is measured for reasons other than meeting the requirements I of specification 4.2.2.2 an overall measured F (z) shall be obtained from a power
- 9 distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5%* to account for measurement uncertainty, f ,
t i
-t 9
*For Unit 1, Cycle 7, when the number of available moveable detector thimbles h is greater than or equal to 50% and less than 75% of the total, the 5% mea- s surement' uncertainty shall be increased to [5% + (3-T/14.5)(2%)] where T ]
is the-number of available thimbles. i
.McGUIRE - UNITS 1 and 2. 3/4{t-9b f(Unit 1)- '
Amendment Amendment-No.A . No.1}VUnit 2
_ = - - _ _ - . - - . _ . _ - . - . - - _ - _ - - _ _ - _ . . - - _ .- .- _ _ _. POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation specified in the CORE OPERATING LIMITS REPORT (COLR) for four loop operation:
. Where:
N
- a. R=
- RTP [1. 0 MFAH (1.0 - P)]
bH THERMAL POWER ,
- b. P
= RATED THERMAL POWER
- c. F H=MeasuredvaluesofF$g obtained by using the movable incore detectors to obtain a power distribution map. The measured values of F H shall be used to calculate R since the figure specified in the COLR includes penalties for undetected feedwater venturi fouling of 0.1% and for measurement uncertainties of 1.7% for flow and 4%* for incore measurement I of F g,
- d. F RTP= The F h limit at RATED THERMAL POWER (RTP) specified in the COLR, and
- e. MF3g= The po.ver factor multiplier specified in the COLR.
APPLICABILITY: ACTION: MODE 1. [/)A>TT 2-) With the combination of RCS total flow rate and R outside the region of acceptable operation specified in the COLR: E. Within 2 hours either:
- 1. Restore the combination of RCS total flow rate and R to within the above limits, or P. keduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
*For Unit 1, Cycle 7, when the number of available moveable detector thimbles g is greater than or equal to 50% and less than 75% of the total, thRTp mea- 9 surement uncertainty shall be increased by changing the value of f in the R equation to [(0.0149/14.5)T + 1.4453) where T is the number N available 3 thimbles. -
McGUIRE - UNITS 1 and 2 3/4 />2-14 AmendmentNo.lkI(Unit 1) Amendment No AiKunit 2)
lLI S S POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued) b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours,
- c. Identify and correct _the cause of the out-of-limit condition prior to increasing-THERMAL POWER above the reduced THERMAL POWER limit '
required by ACTION a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of R and-indicated RCS total flow rate are demonstrated, through incore flux mapping and s RCS total flow rate comparison, to be within the region of acceptable operation specified in the COLR prior + meteding the following THERMAL POWER 1evels: .( '
- 1. A nominal 50% of RATED THERMAL POWER, '
l
- 2. A nominal 75% of RATED THERMAL POWER, and
- 3. Within 24 hours of attaining greater than or equal to 95% of !
RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The combination of indicated RCS total flow rate determined by process computer readings or digital voltmeter measurement and R shall be within the region of acceptable operation specified in the COLR: 3
- a. -Prior to operation above 75% of-RATED THERMAL POWER after each fuel loading, and
- b. At least once per 31 Effective Full Power Days.
4.2.3.3 The indicated RCS total flow rate shall be verified to be within the
- region of acceptable operation specified in the COLR at least once per 12 hours -l '
when the most recently obtained value of R obtained per Specification 4.2.3.2, 4 is assumed-to-exist. 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. , 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. i McGUIRE _ UNITS 1 and 2 3/4h2-15 Amendment No@)l(Unit 1) AmendmentNo.gunit2) u .; . _ _ _ _ . _ _ _ _ _ . . . . . _ _ _ _ , .
() - ta' All / C1 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ([M dl I 1 O b)L ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Within 2 hours either:
a) Reduce the OUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
- 3. Verify that the QUADRANT POWI.1 TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
- 4. Identify and correct the cause of the out-of-liniit condition prior to increasing THERMAL POWER; subsequent P0t'R OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%
or greater RATED THERMAL POWER.
*5ee Special Test Exception 3.10.2.
I h McGUIRE - UNITS 1 and 2 3/4[b-19 Amendment No. (Unit 1) k Amendment No.J (Unit 2) . 1
t)W ' POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION _, ACT l_0_N : (Continued) (
- b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes;
- 3. Verify thet the QUADRANT POWER TILT RATIO is within its limit within 2 hourt after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to $5% of RATED THERMAL POWER within the next 4 hours; and
- 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%
or greater RATED THERMAL POWER.
- c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. McGUIRE - UNITS 1 and 2 3/4f2-20 Amendment No.3 (Unit 1) [7 Amendment No. (Unit 2)
( { POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
- 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER.
?
SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and
- b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. I McGUIRE - UNITS 1 and 2 3/4&2-21 Amendment No. Unit 1) Amendment Nv. (Unit 2)
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS Yl[ S LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1b (uvrf L)
- a. Reactor Coolant System T,yg, and
- b. Pressurizer Pressure.
APPLICABILITY: MODE 1[ ChVTI [ ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of ' RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 shall be measured by averaging the indications (meter or computer) of the operable channels and verified to be within their limits at least once per 12 hours. McGUIRE - UNITS 1 and 2 3/462-22 Amendment No. Unit 1) Amendment No. (Unit 2) m _ - _ . . . _ _ _ . _ . . . . . , _ , _ _ . , , _ , _ . _ . . . . , , , _ , , - . . _ . . ,_.g,, - - , _ . . . - , - - _ . _ _ - , , . . _ - . . .
OlY l TABLE 3.2-lb DNB PARAMETERS
# Oh :ABLE PARAMETER INDICATION CHANNELS LIMITS
- Indicated Reactor Coolant System T meter 4 <590.5 F
~
8V9 meter 3 590.2 F
;9mnuter 4 c591.0 F computer 3 [590.8F Indicated Pressurizer Pressure ** meter 4 ;2226.5 psig meter 3 32229.8 psig computer 4 32221.7 p;ig computer 3 ;2224.2 psig l
l
)
i l l 1 , d
- Limits applicable during four-loop operation. t
** Limits.not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%
RATED THERMAL POWER. McGUIRE - UNITS 1 and 2 - 3/4h-23 AmendmentNo.jJ'(Unit 1) Amendment NoA (Unit 2)
/M Z ;g. TABLE 3.3-1b' . ng 5 REACTOR TRIP SYSTEM INSTRUMENTATION R_ MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE -c ~$ . FUNCTIONAL UNIT. OF CHANNELS. TO TRIP OPERA 8LE MODES ' ACTION d -
2- 1 g 1. Manual Reactor Trip 2 '1 1, 2
,
- 2 1 2 3 * , 4 *, 5* 10 g '. -
m 2. Power Range, ' Neutron Flux -' High 4 2 3 1, 2 2 b Setpoint low 4 2' 3 1,,, , 2 2. '[ Setpoint
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 i High Positive Rate
- 4 Power Range, Neutron Flux, 4 2 3 1, 2 2 9 h High Negative Rate Intermediate Range, Neutron Flux 2 1 2 1 ,2 3 5.
- 6. Source Range, Neutron Flux 2,,
-2 1 2 4
- a. ;Startup
- b. Shutdown 2 1 2 3*, 4*, 5* 10
- c. Shutdown 2 0 1 3, 4, and 5 5
.g R 7. Overtemperature AT.
'S 6 P c' - Four Loop Operation 4 2 '3 1, 2 44 Three Loop Operation (**) (**) (**) (**) (**)
@y n 5;; 30
UnrT L i (
- x. - TABLE 3.3-2b !
t n= , E REACTOR TRIP. SYSTEM INSTRUMENTATION RESPONSE TIMES -[ I E m FUNCTIONAL UNIT- RESPONSE TIME c. I b Manual Reactor Trip g - 1. N . A. t b F [ 2. ' Power Range, Neutron Flux 10.5 second (1) {; o
- 3. Power Range, Neutron Flux, High Positive Rate N. A.
! 4. Power Range, Neutron Flux, j :- High Negative Rate 10.5 second (1) p. ! 5. Intermediate Range, Neutron Flux N.A. [' w. s A 6. Source Range,. Neutron Flux N. A.' l 3 . E 7. Overtemperature AT 110.0 seconds (1)(2)C31 I
- 8. Overpower AT $10.0 seconds (1)(2)(M $
L
.9. Pressurizer Pressure--Low 12.0 seconds >> 10. Pressurizer Pressure--High- <2.0 seconds i ea 2 ee EE 11. Pressurizer Water Level--High N.A.
i ! 2.
- SS i EE (1)' Neutron detectors are exempt from response. time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
$$ (2) The i 10.0 second response time includes a 6.5 second delay for the RTDs mounted in thermowells. g j nm (31'T W W O s M and r hb t~eecktmit-enly4th it: !D byp= =i fem iw h'cd ; un*J1-ther t % + < --R 0 -sec. 3 ' E E- ~'
c - \;X b
. .. . - ._ -. _n. . . . ,
it/IY . 1 ABLE 4. 3-l b 1
~
h REACIOR IRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E m TRIP
- ANALOG ACTUATING MODES FOR c CHANNEL DEVICE WHICH E CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE N FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 1. Manual Reactor Trip N.A. N.A. N.A. R (11) N.A. 1, 2 , 3 * , 4 * , $
- i m 2. Power Range, Neutron Flux High Setpoint S O(2, 4), M N.A. N.A. 1, 2 M(3, 4),
Q(4, 6), ! R(4, 5) j Low Setpoint 5 R(4) M N.A. H.A. 1,,,, 2 U 3. Power Range, Neutron Flux. N.A. R(4) M N.A. N.A. 1, 2 High Positive Rate h. U 4. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 liigh Negative Rate
- 5. Intermediate Range, S R(4, 5) S/U(1),M N.A. N.A. l ,2 Neutron Flux
- 6. Source Range, Neutron Flux S R(4, 5) S/U(1),M(9) N.A. N.A. 2 , 3, 4, 5
- 7. Overtemperature ai S R H N.A. N.A. 1, 2
(( aa
- 8. Overpower AT S R H N.A. N.A. 1, 2
{% EE 9. Pressurizer Pressure--Low S R H N.A. N.A. 1 1,, L
- 10. Pressurizer Pressure--High S R M N.A. N.A. 1, 2
$$ 11. Pressurizer Water Level--High S R H N.A. N.A. I ro r ~
- 12. Low Reactor Coolant F lew 5 R H N.A. N.A. I
^
, /)) *[~ 2..
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION , 3.3.2 The Engineered Safety Features Actuation Systec (ESFAS) Instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip i Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4La t ESPONSE TIMES as shown in Table 3.3-54(dgy g APPLICABILITY: As shown in Table 3.3-3. Q(gl)\T 2 O R)L-l) ACTION:
- a. With an ESFA5 Instrumentation channel or interlock Trip Setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the Trip Setpoint adjusted consistent with the Trip Setpoint value.
- b. With an ESFAS Instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMEN'S 4.3.2.1 Each ESFAS Instrumentation. channel and interlock and the automatic actuation logic and-relays shall be demonstrated OPERABLE by the performance of the E5FAS Instrumentation Surveillance Requirements specified in Table 4.3-2. 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels
- are tested at least once per N times 18 months where N is the total number of
. redundant channels in a specific ESFAS function as shown in the " Total No. of .
Channels" column"bf Table 3.3-3.
'L l
McGUIRE - UNITS'1 and 2 3/4 63-15 i _ _ . . . . . . __.,__,1. _.._,,..,.,__,._.._,....,,_.........,u__.m__._, . _ . . _ _ . . _ . , , _ , , , , .
TABLE 3.3-46
? ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
[]
"o m
TRIP SETPOINT ALLOWABLE VALUES
- FUNCTIONAL UNIT C
5E 1. Safety Injection, Reactor Trip, gt Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and
, Nuclear Service Water.
E N.A. N.A. n,
- a. Manual Initiation Automatic Actuation Logic N.A. N.A.
b. and Actuation Relays
- c. Containment Pressure--High 5 1.1 psig 1 1.2 psig
- d. Pressurizer Pressure--Low-Low 1 1845 psig 1 1835 psig i ~~
Steam Line Pressure - Low 1 565 psig h e. 1585psig U$ 2. Containment Spray 7 N.A. N.A.
- a. Manual Initiation Automatic Actuation Logic N.A. N.A.
- b. -
and Actuation Relays
- c. Containment Pressure--High-High 1 2.9 psig i 3.0 psig
~
a
\
N
YUW ~2.- TABLE 3.3-4b(Continued) W
- g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SElPOINTS m
i-
- FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- c
-v$, 3. Containment Isolation
- a. Phase "A" Isolation i
a 1) Manual Initiation N.A. N.A. m 2)' Automatic Actuation Logic N.A. N.A. and Actuation Relays
- 3) Safety Injection See Item 1. above for all Safety Injection Tria Setpoints and Allowable Values w b. Phase "B" Isolation
- 1) Manual Initiation N.A. g N.A.
- 2) Automatic Actuation Logic N.A. H.A.
and Actuation Relays
- 3) Containment Pressure--High-High 1 2.9 psig 1 3.0 psig
- c. Purge and Exhaust Isolation i
- 1) Manual Initiation N.A. N.A.
- 2) Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- 3) Safety Injection See Iterr 1. above for all Safety Injection Trip Setpoints and Allowable Values l
b/A lI L l Ta9LE3.3-4b(Continued) O ENGINEERED SAFETY- FEATURES ACTUATION SYSTEM INSTRUMENT ATION TRIP SETPOINTS 5 m TRIP SETPOINT ALLOWABLE VALUES [ FUNCTIONAL UNIT
- 14. Steam Line Isolation
- H.A. N.A.
p a. Manual Initiation to Automatic Actuation Logic N.A. H.A. b. and Actuation Relays Containment Pressure--High-High 5 2.9 psig 5 3.0 psig c.
$ 120 psi with a j
- d. Negative Steam Line $ 100 psi with a Pressure Rate - High rate / lag function rate / lag function m time constant time constant
} 1 50 seconds 1 50 seconds 3
1 585 psig 1 565 psig
- e. Steam Line Pressure - Low
- 5. Turbine Trip and reedwater Isolation N.A. N. A.
- a. Automatic Actuation Logic and Actuation Relays
$ 83% of narrow range j ggg
(( b. Steam Generator Water level-- High-High (P-14)
< 82% of narrow range instrument span each steam instrument span each steam generator
)
generator ggj A5 c. ' Doghouse Water Level-High 12" 13" _. z f ,o (Feedwater Isolation Only)
- 6. Containment Pressure Control System CC 0.25 5 SP/T i 0 45 PSIG Start Permissive / Termination 0.3 1 SP/T $ 0.4 PSIG hh (SP/T) mg
. . - ..1.
._ i
/J W 7, ON4E pS GbJrr /
a TABLE 3.3-4(Continuedl ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E m FUNCTIONAL UNIT TRIP SEIP0lNT ALLOWABLE VALUES
] 7. Auxiliary feedwater e a. Manual Initiation N. A. N.A.
E b. Automatic Actuation Logic N.A. N.A. [ and Actuation Relays
- c. Steam Generator Water Level--Low-Low
- 1) Start Motor-Driven Pumps > 12% of span from 0 to > 11% of span from 0 to 50% of RATED THERMAt POWER, 30% of RATED THERMAL POWER, increasing linearly to increasing linearly to M > 40.0% of span at 100% > 39.0% of span at 100%
of RATED THERMAL POWER. h 5f RATED THERMAL POWER. r'o
- 2) Start Turbine-Driven Pumps >12% of span from 0 to > 11% of span from 0 to 30% of RATED THERMAL POWER, 30% of RATED THERMAL POWER, increasing linearly to increasing linearly to
> 40.0% of span at 100% > 39.0% of span at 100%
if RATED THERMAL POWER. of RATED THERMAL POWER.
- d. Auxiliary feedwater -> 2 psig -> 1 psig EE Suction Pressure - Low
@$ (Suction Supply Automatic gg Realignment) 55 e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints z2 Start Motor-Driven Pumps and Allowable Values oo
> 3200 volts
- f. Station Blackout - Start 3464 1 173 volts with a Motor-Driven Pumps and 8.5 1 0.5 second time 22 lurbine-Driven Pump delay E. 3. (Note 1) . 3 or 'l*
w g. Trip of Main feedwater Pumps - N.A. N.A. Start Motor-Driven Pumps .
d/Td SW4E As Outt q TABLE 3.3-46(Continued)
?
@ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m
' TRIP SETPOINI ALLOWABLE VALUES FUNCTIONAL UNIT
- 8. Automatic Switchover to Recirculation
]
~ RWST Level > 90 inches 3 80 inches
't
" 9. Loss of Power m
4 kV Emergency Bus Undervoltage- 3464 1 173 volts with a 3 3200 volts Grid Degraded Voltage 8.5 1 0.5 second time delay
- 10. Engineered Safety features Actuation System Interlocks Pressurizer Pressure, P-Il < 1955 psig 5 1965 psig t' a.
$ 3 553 F 3 551"F
- b. T avg, P- 2
{ e 'N. A.
- c. Reactor Trip, P-4 N.A.
Steam Generator Level, P-14 See item S. above for all Trip Setpoints and Allowable l d. Values. e NN lhe turbine driven pump will not start on a blackout signal coincident with a safety injection signal. 6 l
@@ Note 1:
aa F. F 22 - S. 3. ne
l TABLE 3.3-5 /U /T 1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. M nual
- a. Safety Injection (ECCS) N.A.
- b. -Containment Spray N.A.
- c. Containment Isolation Phase "A" Isolation N. A.
Phase "B" Isolation N. A. Purge and Exhaust Isolation N. A.
- d. Steam Line Isolation N. A.
- e. Feedwater Isolation N.A.
- f. Auxiliary Feedwater N.A.
- g. Nuclear Service Water N.A.
- h. Component Cooling Water N.A.
- i. keactor Trip (from SI) N.A.
J. Start Diesel Generators N.A.
- 2. Containment Pressure-High
- a. Safety Injection (ECCS) 1 27(1)
- b. Reactor Trip (from SI) 12
- c. Feedwater Isolation <9
~
- d. Containment Isolation-Phase "A"(2) 5 18(3)/28(4)
- e. Containment Purge and Exhaust Isolation 54 l
- f. Auxiliary Feedwater(5) N.A.
- g. Nuclear Service Water 1 65(3)/76(4)
- h. Component Cooling Water 1 65(3)/76(4)
- i. Start Diesel Generators 5 11 McGUIRE - UNITS 1 & 2 3/4!G-30 Amendment No.1% (Unit 1)
Amendment No49 (Unit 2) 2/40/90
U m 7 2-l TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 3. Dressurizer Pressure-Low-Low
- a. Safety Injection (ECCS) 1 27(1)/12(3)
- b. Reactor Trip (from SI) 12
- c. Feedwater Isolation <9
- d. Containment Isolation-Phase "A"(2) 18(3)/28(4)
- e. Containment Purge and Exhaust Isolation 54 l
- f. Auxiliary Feedwater(5) N.A.
- g. Nuclear Service Water System 5 76(1)/65(3)
- h. Component Cooling Water 1 76(1)/65(3)
- 1. Start Diesel Generators 1 11
- 4. Steam Line Pressure-Low
- a. Safety Injection (ECCS) 1 12(3)/22(4)
, b. Reactor Trip (from SI) $2
- c. Feedwater Isolation <9
- d. Containment Isolation-Phase "A"(2) 1 18(3)/28(4)
- e. Containment Purge and Exhaust Isolation 54 l
- f. Auxiliary Feedwater(5) N.A.
- g. Nuclear Service Water 1 65(3)j7g(4)
- h. Steam Line Isolation <7
- 1. Component Cooling Water 65(3)/76(4)
- j. Start Diesel Generators 1 11
- 5. Containment Pressure-High-High
- a. Containment Spray 1 45
- b. Containment Isolation-Phase "B" N.A.
- c. Steam Line Isolation 17
- 6. Steam Generator Water level-High-High
- a. Turbine Trip N.A.
- b. Feedwater Isolation 59 l McGUIRE - UNITS 1 & 2 3/463-31 Amendment No. %QI(Unit 1) '
l Amendment No.44 (Unit 2) 2ftet90
(J IT TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 7. Steam Generator Water Level LowLow
- a. Motor-driven Auxiliary Feedwater Pumps 1 60
- b. Turbine-driven Auxiliary Feedwater Pumps 1 60
- 8. Negative Steam Line Pressure Rate - High l Steam Line Isolation 57 l
- 9. Start Permissive l Containment Pressure Control System N.A. I
- 10. Termination l
! Containment Pressure Control System N.A. , 1
- 11. Auxiliary Feedwater Suction Pressure - Low I Auxiliary Feedwater Pumps (Suction l l- Supply Automatic Realigrment) 1 13 l l
12, RWST Level ' Automatic Switchover to Recirculation 1 60
- 13. Station Blackout
- a. Start Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Start Turbine-Driven Auxiliary Feedwater Pump (6) .1 60
- 14. Trip of Main Feedwater Pumos Start Motor-Driven Auxiliary Feedwater Pumps 1 60
- 15. Loss of Power 4 kV Emergency Bus Undervoltage- ~
< 11 I
Grid Degraded Voltage
+e..
! McGUTRE - UNITS 1 and 2 3/4['3-32 Amendment No. (Unit 2) Amendment No (Unit 1) l l
[jp l REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVEj N SHUTOOWN - LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lif t setting of 2485 psig i 1%.* APPLICABILITY: MODES 4 and 5(()A)TT L CAJL ACTION: With no pressurizer Code safety valve OPERABLE, imm loop into operation in the shutdown cooling mode. SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by Specification 4.0.5.
*The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
3/484-7 McGUIRE - UNITS 1 and 2
REACTOR COOLANT SYSTEM i ! O /JI T 2' OPERATING LIMITINGJNDITIONFOROPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig .K ^ _ g e,b APPLICABILITY: MODES 1, 2, and 3.h// Jf T 2-ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable salve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTOOWN within the following 6 hourr. SURVEILLANCE FEQUIREMENTS 4.4.2.2 No adcitional requirements other than those required by Specification 4.0.5.
*The lift setting pressure shfl correspond to ambient conditions of the valve at nominal operating temperature and pressure.
McGUIRE 'JNITS 1 and 2 3/tJ4-8
- i. II b 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYC E SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lif t settings as specified in Table 3.7-3f, Mr))'J~Jh APPLICABILITY _: MODES 1, 2, and 3.[()AJi { Z.).
ACTION:
- a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Tabic 3.7-1; otherwise, be-in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With three reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. McGUIRE - UNITS 1 and 2 3/487-1 l
- - - , , . . . - - .< ~ -
YA)f [' -
.x
-TABLE 3.7-3h
; h i
.I m STEAM LINE SAFETY VALVES PER LOOP
. VALVE NUMBER LIFT SETTING (+ 1%)* ORIFICE SIZE Loop A Loop B Loop'C ' Loop D f [
2
- 1. SV 20 SV 14 SV 8 SV 2 1170 psig 12.174 in ,
j 2
- 2. SV 21 SV 15 SV 9 SV 3 1190 psig 12.174 in 2
- 3. SV 22 SV 16 SV 10 SV 4 1205 psig 16.00 in
- 4. SV 23 SV 17 SV 11 SV 5 1220 psig 16.00 in 2
- 5. SV 24 SV 18 SV 12 SV 6 1225 psig 16.00 in 2 D{
i I Ln2 t a
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
O lL)) PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.4 Each main steam line isolation valve (MSLIV) shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3.(/ OA /T 2 CA>L@ ACTION: MODE 1 - With one MSLIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours. MODES 2 - With one MSLIV inoperable, subsequent operation in MODE 2 or 3 may and 3 proceed provided:
- a. The isolation valve is maintained closed, and
- b. The provisions of Specification 3.0.4 are not applicable.
Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 Each MSLIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. McGUIRE - UNITS 1 and 2 3/8-8
l 2.1 SAFETY LIMITS YA)lT R l l BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above'the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR), defined as the ratio of the heat flLx that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The DNB design basis is as follows: there must be at least a 95% proba-bility that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR set such that there is a 95% probability with 95% confidence that DNB will not occur when the minimum DNBR is at the DNBR lirait. In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertain-ties in the above plant parameters are used to determine the plant DNBR uncer-tainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. The curves are based on a nuclear enthalpy rise hot channel factor, FNg, of 1.49 and a reference cosine with a peak of 1.55 for axial power shape. An allow-ance is included for an increase in F H at reduced power based on the expression: F"g = 1_49 [1 + 0.3 (1-P)] Where P is the fraction of RATED THERMAL POWER. McGUIRE - UNITS 1 and 2 B62-1 Amendment No. [(Unit 1) Amendment No (Unit 2)
()UIT 'l LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued) The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. The Power Range Negative Rate trip provides-protection for control rod drop accidents. At high power, a rod drop accident of a single or multiple rods could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBR's will be greater than-the design limit DNBR value. , Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences 'of an uncon-trolled rod cluster control assembly bank withdrawal-from a suberitical ! condition. These trips provide redundant protection to-the low Setpoint trip i of the Power Range, Neutron Flux chgnnels. The Source Range channels will l initiate a Reactor trip at about 10 5 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. 1 i McGUIRE - UNITS 1 and 2 882-4 AmendmentNo.h(Unit 1) Amendment No AS (Unit 2)
- _- -- . - -- . . - . ~ -. . -
0)(T 2-3/4.2 POWER OISTRIBUTION LIM 1T$ BASES The specifications of this section provide assurance of fuel integrity during events Condition _I (Normal Operation) and II (Incidents of Moderate Frequency) by: (1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in_short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop-erties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. The-definitions of certain hot channel and peaking factors as used in these specifications are as follows: F0 (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing toler-ances on fuel pellets and rods; F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX OIFFERENCE (AFO) assure that the F (Z) upper 9 bound envelope of the-F limit specified in the CORE OPERATING LIMITS REPORT x (COLR) times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position- for steady state operation at high power levels. The value of the-target flux difference obtained under these conditions divided by the fraction 7 of RATED THERMAL POWER is the target flux difference _at RATED THERMAL' POWER for the associated core burnup conditions. Target flux differences for other
. THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER va?ue-by the appropriate fractional THERMAL POWER-level. The periodic updating of the target flux difference value is necessary to' reflect core burnup considerations.
McGUIRE - UNITS 1 and 2 B3/4E2-1 Amendment No.1 (Unit 1) Amendment No. 7(Unit 2)
l l POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) At power levels below APL"0, the limits on AFD are defined in the COLR, i.e. that defined by the RAOC operating procedure and limits. These limits were l calculated in a manner such that expected operational transients, e.g. load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in significant xenon redistribution such that the envelope of peaking factor 5D would change sufficiently to prevent operation in the vicinity of the APL power level. At power levels greater than APL ND , two modes of operation are permissible;
- 1) RA00, the AFD limits of which are defined in the COLR, and 2) base load %
operation, which is defined as the maintenance of the AFD within a COLR specified band about a target value. The RA0C operating procedure above APL ND > is the same as that defined for operation below APLNO However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with Fq (2) less than its limiting value. To allow operation at the maximum permissible value, the base load operating procedure restricts the indicated AFD to relatively small target bcnd and power swings (AFD target band as specified in the COLR, APLND < power < APLBL or 100% Rated Thermal Power, whichever is lower). For base load operation, it is expected that the plant will operate within the target band. Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon redistri-bution impact from past operation on the base load operation, a 24 hour waiting period at a power level above Af
- and allowed by RAOC is necessary. During this time period load changes and rod motion are restricted to that allowed by the base load procedure. After the waiting period extended base load operation is p&rmissible.
The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are: 1) outside the allowed al power operating space (for RAOC operation), or 2) outside the allowed AI target band (for base load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation), or 2) APL"O (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed. McGUIRE - UNITS 1 and 2 B 3/402-2 Amendment No.15(Unit 1) l Amendment No. 7(Unit ?) i
h.ll b POWER'0ISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT-FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS accep-tance criteria limit. These limits are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. Each of these.is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position;
- b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and '
- d. The ax?al power distribution, expressed in terms of AXIAL FLUX DIFFEREdCE, is maintained within'the limits, Fhwillbemaintainedwithinitslimits;providedConditionsa,throughd.
- above are maintained. As noted on the figure specified in the CORE OPERATING LIMITS REPORT-(COLR) RCS flow rate and power may be " traded off" against one
- another (i.e. , a low masured RCS flow rate is_ acceptable if the power level is decreased) to ensure t5at th*.-calculated DNBR will not be below the design DNBR value.TherelaxationofFhasafunctionofTHERNALPOWERallowschangesin the radial power shape for.all permissible rod insertion-limits.
R as calculated in Specification-3.2.3 and used in the-figure specified in the COLR, accounts for_ Fh less than or equal to the F limit specified in the COLR. ThisvalueisusedinthevariousaccidentanalyseswhereFh
' influences parameters. other than DNBR, e.g. .. peak clad temperature, and thus is the maximum "as measured" value-allowed.
Margin between the safety analysis limit DNBRs and the design limit DNBRs l is maintained. A-fraction of this margin is utilized to accommodate _the transi-tion core DNBR penalty (2%) and the appropriate fuel rod bow DNBR penalty (WCAP.- 8691,LRev. 1). When an qF . measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate McGUIRE - UNITS 1 and 2 8 3/482-2a- Amendment No.TO3(Unit 1) AmendmentNo./J(Unit 2)
1 POWER 01STRIBUTION L HITS W5 i BASES-i HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RIS HOT CHANNEL FACTOR (Continued) i for a full-core map taken with the Incore Detector Flux Mapping System, and a i 3% allowance is appropriate for manufacturing tolerance. When RCS flow rate and FH are measured, no additional allowances are necessary prior to comparison with the limits of the figure specified in the COLR, Q , Measurement errors of 1.7% for RCS total flow rate and 4% for FN have been allowed for in determination of the design DNBR value. AH The measurement error for RCS total flow rate is based upon performing a
~
precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling 'f j the feedwater venturi is included in the figure specified in the COLR. Any 4 fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending-various plant performance parameters. If detected, action shall'be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation specified on the figure specified in the COLR. f The hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power factor appropriate to either RAOC or base load operation, W(2) or W(z)BL, to provide assurance that the limit on the hot channel. factor,'Fq (z), is met. W(z) accounts for the effects of normal operation transients and was determined from expected power control maneuvers over the-full range of burnup conditions in_the core. W(z)BL accounts for the more restrictive operating limits allowed by base load operation which result in less severe transient values. The W(z) function for normal operation , and the W(z)gg function for base load operation are _specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. Q ' \ i l McGUIRE - UNITS 1 and 2 83/442-4 AmendmentNo,kO5(Unit 1) AmendmentNo.A(Unit 2)
h] l[ [ POWER DISTRIBUTION LIM,ITS BASES 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is prov.ded to allow identification and correc-tion of a dropped or misaligned rod. In the event such action does not cor-rect the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3% f rom RATED THERMAL POWER for each percent of tilt in excess of 1.0. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-ll, N-8. 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the para-meters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are censistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design limit ONBR throughout each analyzed transient. The indicated T 3 values .and the indicated pressurizer pressure ~ values j correspond to analytical limits of 592.6 F and 2220 psia respectively, with 3 allowance for indication irstrumentation measurement uncertainty. j The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Indication pp" ' instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1. \ i I McGUIRE - UNITS 1 and 2 B 3/4f2-5 Amendment No. / (Unit 1) Amendment No 6 (Unit 2)
3/4.4 REACTOR COOLANT SYSTEM Q 1 BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design limit during all normal operations and antici-pated transients. In MODES 1 and 2 with one reactor coolant loop not in oper-ation this specification requires that the plant be in at least HOT STANDBY within I hour. In MODE 3, two reactor coolant loops provide sufficient heat removal l capability for removing decay heat; however, single failure considerations require that three loops be OPERABLE. I In MODE 4, and in MODE 5 with reacter coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change ratt associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting a reactor coolant pump with one or more RCS cold legs less than or equal to 300 F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients vd will not exceed the limits of Appendix G by either: (1) restris...ig the water volume in the pressurizer and thereby providing a volume for the reactor coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water tempera-ture of each steam generator is less than 50 F above each of the RCS cold leg temperatures. Amendment No. Unit 1) McGUIRE - UNITS 1 and 2 B 3/4fA-1 Amendment No (Unit 2)
1 l l 1 l l 1 Attachment Ib Technical Justifications
Proposed Revision to Technical Specification Floure 2.1-1 This change applies to Unit 1 only. This proposed Technical Specification (TS) revision changes Figure 2.1-1 to reflect use of the BWCMV CHF correlation and Duke Power Company's Statistical Core' Design (SCD) methodology with a 1.55 thermal design DNBR limit. Technical Justification With the first batch implementation of the Mark-BW fuel design, Duke Power Company (DPC) has recalculated the McGuire reactor core-safety limits using the BWCMV CHF correlation (Reference 1) along with its Statistical Core Design (SCD) methodology, Reference 5. With the: implementation of these design methodologies, it was possible to slightly increase the nuclear enthalpy rise hot channel factor, F$i, from 1. 4 9 to 1. 50, to allow greater fuel cycle design flexibility. The proposed changes to Figure 2.1-1 reflect the use of this new design limit as well as the use of BWCMV and SCD. The reactor core safety limits provided on Figure 2.1-1 depict i the combinations of thermal power, reactor coolant system pressure, and a verage temperature within which the calcul.ated DNBR is no less than the design limit DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid. The analysis which defines these limits is based on a full core of Mark-BW assemblies with a thermal design flow rate that bounds the minimum measured flow at McGuire. The DNB limited portions of these curves are defined using the BWCMV CHF correlation with'a design DNBR limit of 1.55. This design DNBR limit provides 10.7 percent thermal margin to the 1.40 BWCMV statistical design limit which is defined for the McGuire core using the DPC SCD methodology. The safety limits are based on a design peaking distribution with a nuclear enthalpy rise hot channel factor, FU,, of 1.50 and a reference chopped cosine axial power shape with a peak of 1.55- To verify that this design peaking distribution is conservativ on a cycle-specific basis, maximum allowable peaking (MAP) limits, which provide DNB equivalence to the design distribution at various safety limit statepoints, are defined. To verify that margin was available, _these MAP limits were compared to predicted cycle-specific peaking distributions, Reference 2. As part of the safety limit / MAP limit analyses, an evaluation was performed which showed that, if power was reduced below 100 percent, peaking could be increased according to the following relationship: k=1 + 0.3(1 - P) Ib-2
where k = the factor by which the MAP limits are adjusted to define reduced power limits P = the fraction of rated power while maintaining the core within the aforementioned thermal limits. Comparison of the Mark-BW safety limits and the Westinghouse OFA l safety limits shows that at all points the Mark-BW litaits are outside the OFA limits. Mixed core studies have shown that the Mark-BW safety limits are applicable to the Westinghouse OFA fuel if a DNBR penalty is included for those assemblies. The DNBR penalty for OFA fuel is applied against the 10 7 % margin included in the design DNBR limit. Eronosed Revision to Technical Snegification Table 2.2-1 1 This change applies to Unit 1 only. This proposed Technical Specification (TS) revision changes the K values for the overtemperature and overpower AT trip functions to i reflect the use of the BWCMV CHF correlation and Duke Power Company's Statistical Core Design (SCD) methodology with a 1.55 I thermal design DNBR limit. In addition, an axial imbalance l penalty, f2 ( AI) , is applied to the OPAT reactor trip. Finally, I I the Power Range Neutron Flux Negative Rate reactor trip is deleted from the Reactor Protection System. Technical Justification l The overtemperature and overpower AT reactor trips are designed to protect-the. reactor from DNBR and centerline fuel melt (CFM).
- 1 Due to the new DNBR methodology, the allowable operating region .is modified. Therefore, new K values are calculated to conservatively bound this operating region.
The purpose of the OPAT trip function is to prevent center-line fuel melt (CFM) during normal operation and condition II transients. The OPAT trip function is designed to trip the reactor when the measured AT exceeds 118% of normal full power AT. _The f2 (AI) portion of the trip function is designed to lower the trip setpoint when axial flux differences (AFDs) exceed predetermined limits. Since the limiting margins to CFM occur as the result of highly skewed power distributions, a f2 (AI) trip reset function can be developed to prevent CFM limits from being exceeded, or increase the available margin to the CFM limit. Analysis of the M1C8 core indicates that no f 2 (AI) trip function Ib-3
l l l is required to protect the core from CFM. However, from an operational and design standpoint, it is desirable to eliminate from consideration power distributions corresponding to high AFDs. For the MIC8 core, the f 2 (AI) trip reset function was developed to produce a reactor trip on high AFDs for credible overpower events protected by the OPAT trip function. The impicmente. tion of this function eliminates from consideration power distributions for highly skewed conditions and affords incrsased margin to the CFM limit. This margin is important in that it will reduce the probability that adjustments to the OTAT trip function will be required as the result of CFM surveillance requirements. A reactor trip on negative flux rate is not assumed in any of the licensing basis accident analyses. The analysis of the dropped rod accident (Reference 4), for which the negative flux rate trip was designed to provide protection, assumes that reactor trip occurs on low pressurizer pressure, if at all. For cases in which no trip occurs, i.e. for low dropped rod worths, this analysis shows that none is needed. The removal of the negative flux rate trip from the reactor protection system will eliminate unnecessary _ reactor trips resulting from such low worth rod drop events. Proposed Revision to Technical Specification Table 3.1-1 This change applies to Units 1 and 2. The table is revised to include all accident analyses that would require reevaluation in the event that one full-length Rod Cluster Control Assembly is inoperable. Technigal Justification The existing Table 3.1-1 lists the Rod Cluster Control Assembly Misalignment event-when, in fact, the broader-scoped event of RCCA Misoperation should have been given. This accident includes the static misalignment as well as single rod withdrawal, dropped rod and dropped bank events, which might all be impacted by the inoperable rod. The large break LOCA analysis, which was listed in the table, does not take credit for any control rod insertion and should therefore be removed. Proposed RevisiSn to Technical Specifications 3/4.2.1, 3/4.2.2. j 3/4.2.3. 3/4.2.4, and 3/4.2.5 The current Power Distribution TS have been changed to reflect applicability to Unit 2 only. This has been done by placing i "(Unit 2)" in the Applicability Section of the current TS. The Ib-4 i
\
Unit 2 Power Distribution TS will have "B" placed in front of the page no'ber (Ex: 3/4 B2-7b), and the Unit 1 Power Distribution chang < ill have "A" placed in front of the page numbers. The l Unit 4 , will be changed as marked up and copied on white paper. , The Ur., . 2 Power Distribution TS will be changed as described and I copied on yellow paper. Proposed Revision to Technical Specification 3/4.2.1 This change applies to Unit 1 only. The target AFD for Base Load operation and the RAOC limits have been replaced with an envelope of allowed AFD values at various power levels. The AFD setpoints given in the COLR replace the RAOC operating space referred to in the current Specification. Since the reactor is not constrained to operate at a specified target AFD, the target AFD and associated band have been eliminated from Specification 3.2.1. The allowable operating AFD space is anywhere within the AFD setpoint envelope in the COLR. Technical Justification Specification 3.2.1 was revised to provide an LCO statement and required actions consistent with Duke Power Company methodology for core power distribution control as discussed in DPC-NE-2011PA (Reference 2). During the-reload licensing analyses for McGuire Unit 1, Cycle 8, a three-dimensional maneuvering analysis was performed to determine the Axial Flux Difference (AFD) limits based on the methodology of DPC-NE-2011PA. Specification 3/4.2.1 was revised for consistency with the new analytical methodology and to reflect the results of the cycle 8 analysis. The resulting AFD setpoints were placed in the Core Operating Limits Report (COLR). The AFD limits of Specification 3.2.1 prevent the_ core power distribution from exceeding the allowable values based on the LOCA peaking limits and the initial condition DNB maximum. allowable peaking (MAP) limits during power operation. . The AFD limits' arc defined by a three-dimensional core maneuvering analysis that determines core peaking dependence on core loading, fuel depletion, thermal-hydraulic statepoint, control rod position, and xenon distribution. Correlations between peaking margin and axial power offset are developed that allow
~
determination of negative and positive offset limits at selected power levels. The resulting offset limits preclude operation with negative margin, and are translated into corresponding AFD limits. The peaking margins are calculated by augmenting nodal peaks with uncertainties and allowances as described in DPC-NE-2011PA. The margin database comprises calculations from the entire range of power distributions generated in the maneuvering analysis, including control bank insertion to the insertion limit Ib-5
l ! and transient xenon conditions. The limiting K(Z) peaking limits corresponding to the composite of the K(Z) limit for Mark-BW and OFA fuel were used in the maneuvering analysis to compute LOCA margin. The AFD limits determined for cycle 8 were adjusted for measurement uncertainty and also include the peaking increase corresponding to a Quadrant Power Tilt ratio of 1.02. The adjustment was applied at each power level between 100% and 50% of rated thermal power. The AFD limits given in the cycle 8 COLR are the adjusted AFD limits. K(Z) limits for the Mark-BW fuel were determined by the ECCS analysis performed for Mark-BW fuel, as documented in BAW-10174 (Reference 3). The K(Z) limits for the OFA fuel are the current Westinghouse K(Z) values for McGuire, as given in current McGuire COLRs. Initial condition DNB peaking margins were computed from the augmented peaks and the MAP limits based on Statistical Core Design (SCD) methodology, as described in DPC-NE-2004 (Reference 5). The MAP limits are a family of peaking limits for which either the minimum DNBR is equal to the design DNBR limit, or the coolant quality at the minimum DNBR location is equal to the CHF correlation quality limit. The MAP limits provide linkage between the reference DNBR analyses, with their design peaking distributions, and the core operating limits. The initial condition Map limits are based on the statopoint that represents the point of minimum DNBR during the most limiting non-OTAT DNB transient. The initial condition MAP limits are based on a DNBR limit that includes 10.7% margin which is more than enough to account for the mixed core DNBR penalty for OFA fuel. Proposed Revision to Technical Specification 3/4.2.2 This change applies to Unit 1 only. Specification 3/4.2.2 was revised to reflect the power peaking surveillance method described in DPC-NE-2011PA. These revisions are summarized as follows:
- 1. The statement of the LCO was revised to reflect new nomencloture for the heat flux hot channel factor
[ (F9(.X ,Y, Z) ] required by the methodology in DPC-NE-2011PA and used throughout the Reload Report. Also, as discussed above, separate K(Z) curves are provided for the Mark-BW and OFA fuel types.
- 2. Action a in the current specification has been replaced by Actions a, b, and c in the new specification. The thermal power reduction required when F 9 (X,Y,Z) exceeds its limit Ib-6
m areI the same as the_ current-requirement,-as is the reduction required in the OPAT trip setpoints. Action b is a new requirement, and is provided to limit the allowable AFD when Fn(X,Y,Z) exceeds its limit. This reduces the possibility of-operating _the core in excess of the Fn (X,Y,Z) limit when a margin calculation (discussed in item 7 below) indicates negative-operational margin exists. 3.~- There is no change to SR 4.2.2.1.
- 4. SR 4.2.2.2_ addresses obtaining an incore flux map and the requirements based on the results of the measurement. The reference to RAOC operation has been deleted, since RAOC
-operation is unique to Westinghouse methodology. -5. There is no change to SR 4.2.2.2.a.
- 6. --SR 4.2.2.2.b in-the current surveillance has been deleted.
The-allowances for measurement uncertainty and manufacturing j tolerances have been included in the limit (Fh(X,Y,Z))and therefere the measured peak FU(X,Y,Z) is not increased by these factors.
- 7. SR 4.~2.2.2.c in the current surveillance has been deleted. ;
--No simple determination is made of only whether or not '
the limit has been exceeded.- Instead,-the amount by which the 4.2.2.2 measured value is above or below the limit is
- qualified _as-detailed in' item 10, below.
- 8. SR 4.2.2.2.d:(current surveillance) specifies the frequency for measuring the core power distribution. This is done by part-b'in the -
- new surveillance.
Part b.3 has been added_to this surveillance, requiring an F (X,Y, n Z) measurement when the excore quadrant power tilt ratio is
~
- normalized using incore detector measurements.- This ensures that the . impact of any core _ tilt on nF (X,Y,Z) will be determined'and reflected.in the margin calculations of part M c.-
- 9. iSR 4.2.2.2.e has been replaced by SR 4.2.2.2.d_in the'new surveillance. The intent of these requirements is similar
'insthat' projections of the measurements are made to determine at--whattpoint peaking would exceed allowable limits if the1 current trend continues. -
Inithe new surveillance, an_incore flux map is obtained and a determination is made' as to whether -the measured F (x,y, n z) lwill_ exceed the allowable peaking'at.31 Effective Full-Power Days - (ErPD) -beyond- the :most recent measurement. If the extrapolated F n(x,y,z) measurement ~ exceeds the allowable Fn (x,y, z) l limit, then either the surveillance interval to the next power distribution map is decreased based on the Ib-7
availabic margin, or the F n(x,y, z) measurement is increased by 2% and the margin calculation of 4.2.2.2.c repeated. This surveillance helps ensure that peaking will not exceed allowable limits prior to the next 31 EFPD measurement interval.
- 10. The new SR 4.2.2.2.c replaces 4.2.2.2.f in the current surveillance. The purpose of part col ie to perform margin calculations based on the measured peaks. With the new methodology, the limit ( ( Fh ( X , Y , Z) ) ) to which the moauurement is compared is the design peak at steady-state conditions, increased by a factor that represents the maximum amount that the power at the given assembly location and axial elevation can increase above the design value before the measured value may become limiting. Margins to both the LOCA peaking limit (operational margin) and the centerline fuel melt limit (RPS margin) are calculated. The oporr.tional margin forms the basis for restricting the AFD limits in part c.2, and the RPS margin forms the basis for
'teducing the OTAT trip setpoint in part c.3.
- 31. 3R 4.2.2.2.c.2 (new) replaces SR 4.2.2.2.f.2 in the current surveillance. The reduced AFD limits determined in part c.2 are based on the amount of negative operational margin resulting from the margin calculation of part co l. The negative and positive AFD limits are reduced 1% for each percent change in margin. The AFD must be controlled to those new limits to reduce F n(X,Y,Z) , and to ensure that peaking will be limited for continued power operation.
- 12. SR 4.2.2.2.c.2.b (new) corresponds to SR 4.2.2.2.f.2.b (current surycillance).
- 13. Part 4.2.2.2.c.3 has been added to the surveillance. This part of the surveillance requires reducing the Ki value of the OTAT trip setpoint if the RPS margin is negative. This requirement ensures that centerline fuel melt protection exists when core peaking may be greater than the design values.
- 14. SR 4.2.2.2.f.2.c, which addresses Base Load operation, has been doloted from the new surveillance. The power distribution methodology of DPC-NE-2011PA does not constrain core-operation to a target AFD.
- 15. SR 4.2.2.2.g has been replaced by SR 4.2.2.2.e in the new surveillance; there are no substantive changes to this surveillance.
- 16. SR 4.2.2.3 addresses Base Load Operation and has been deleted from the new surveillance.
Ib-8
- 17. SR:4.-2.2.4 addresses surveillance of peaking in Base Load operation and has been deleted from the new surveillance.
- 13. SR 4.2.2.5 has been replaced by SR 4.2.2.3 in the new surveillance; there are no substantive changes to this surveillance.
l Igghnical Justification Specification 3/4.2.2-was revised to provide required actions and surveillance requirements consistent with Duke Power Company methodology for core power distribution control and surveillance of.-the heat fluxLhot channel factor, as discussed in DPC-NE-2011PA'(Reference 2). lThe heat flux hot channel- factor (F n(X,Y,Z)] -is a specified E ' acceptable fuel design limit that preserves the initial conditions lfor the ECCS analysis. Fn (X,Y, Z) is defined as the maximum local heat flux on the surface of a fuel rod at a given core elevation (Z)- in an assembly located at (X,Y), divided by
-the average fuel rod heat flux, allowing for manufacturing L
tolerances on the fuel pellets and fuel rods. Since F q (X,Y,Z) is a: ratio of local-surface' heat _ fluxes, it is-related to the total local power _~ density inLa fuel rod. Operation within the Fn (X,Y,Z) limits given in the Core Operating Limits Report (COLR) prevents power-peaking-that would exceed the loss of coolant L ' accident.(LOCA)' peaking limits derived by the ECCS analysis. The F n (X, Y, Z) limit is the product.of the peaking limit at rated thermal _ power (F(") and the normalized peaking limit as a function of core elevation (K(Z)]. Limiting K(Z)_ peaking limits basedinn the composite of Mark-BW- fuel and OFA fuel K(Z) peaking p 'limitsiwere :used in the maneuvering' analysis -to determine the operating limits.. The1K(Z) limits for the Mark-BW fuel were determined by the ECCS analyais performed for Mark-BW. fuel, as documented in' BAW-101741(Reference 3) . The K(Z) limits for the OFA-fuelT are the current Westinghouse K(Z) values forLMcGuire, as fgiven in the1 current _McGuire COLR. E 'The: reload maneuvering-analysis defines the axial flux _ difference
-(AFD)cpower level: space,-the rod insertion limits and the;f(AI) penalty-function employed in-the OPAT and/or the OTAT trip functions. Limits-on the-above parameters provide assurance that-core peaking limits are bounded and thermal limits are not challenged. . Appropriate uncertainties as described in DPC-NE-2011PA are applied to calculated: peaks used.to est?blish the'AFD power ~1evel space and the f(AI) penalty function.
Mcasurement ofLthe core power distribution.at' steady-state conditions by using the'incore detectors to obtain a-three-dimensional flux. map provides confirmation that the measured ~ heat Ib-9
l flux hot channel f actor P'(X,Y, q Z) is within tho values of the designed core power distribution. This comparison verifies the applicability of the design power level, control bank insertion, AFD, and excore quadrant power tjlt ratio to the measured core conditions to preserve the LOCA peaking criteria. Proposed Revision to Technical Specification 3/[tita This change applies to Unit 1 only. Specification 3/4.2.3 was revised to reflect the power peaking surveillance method described in DPC-NE-2011PA. These revisions are summarized as follows:
- 1. The statement of the LCO was revised to reflect new nomenclature for the nuclear enthalpy rise hot channel factor (F$n(X,Y)) and related parameters required by the methodology of DPC-NE-2011PA and used throcghout the Reload Report.
- 2. Those requirements of Actions a, b, and c in the current specification relating to the Reactor Coolant System flow rate have been incorporated in Specification 3.2.5. The Actions have been revised to include the reduction of allowable thermal power when Fyn(X,Y) exceeds the limit within 2 hours. The factor (RRH), by which the power level is decreased per percent F3i(X,Y) is above the limit, is defined in the COLR. The inverse of this factor is the fractional increase in the MAPS allowed when thermal power is decreased by 1% RTP. When a power level decrease is required because Pui(X,Y) has exceeded its limit, then Action b requires restoration of Fui(X,Y) to within its limit within 6 hours, or a reduction in the high flux trip setpoint. The amount of reduction of the high flux trip setpoint is governed by the same factor (RRH) that determines the thermal power level reduction. This maintains core protection and an operability margin at the reduced power level similar to that at rated thermal power.
- 3. Action c was replaced by Action d. The partions of the Action requirements related to Reactor Coolant System flow rate have been incorporated in Specification 3.2.5, and do not appear in Action d of the new specification.
- 4. Action item c has been added and requires a reduction in the OTAT K i trip setpoint by an amount equivalent to TRH for each 1% F 3;(x,y) exceeds l'cs limit within 72 hours of initially being outside the limit. This action ensures that the one protection margin is maintained at the reduced power level for DNB related transients not covered by the Ib-10
reduction in the Power Range Neutron Flux-High Trip Sotpoint.
- 5. Thoro is no change to SR 4.2.3.1.
- 6. SR 4.2.3.2 formerly covered only survol11ance frequency. It
, has boon expanded as detailed below to reflect the power peaking surveillance method described in DPC-NE-2031PA and the format of the revised SR 4.2.2.2. Part a addresses obtaining an incore flux map.
- 7. SR 4.2.3.2.b (now) replacos the current 4.2.3.2 and addressos the frequency for confirming that Fu,(x,y) is within its limit. In addition to performing the survol11anco at 1 cast once por 31 EPPD, the revised surveillanco requires measurement of the peaking factor whenover the excore quadrant power tilt ratio is normalized
, using incoro detector measuromonts. This ensures that the i impact of any core tilt on Fui(x,y) will bo datormined and reflected in the margin calculation. This is comparable to the now SR 4.2.2.2.b in the Fn(x,y,z) specification. The survoillance requiring a survoillance to bo performed prior to operation abovo 75% of RATED THERMAL POWER at the bcginning of cach fuel cycle has boon replaced by the requiremont identical to SR 4.2.2.2.b.2 in the Fn(x,y,z) specification. This surveillanco ensures that the plant is at equilibrium conditions prior to a measuromont, and also has a provision similar to the requirement it replaced stating that during power escalation at the beginning of each cycle, THERMAb POWER may be increased until a power lovel for extended operation has boon achieved and a power distribution map obtained.
- 8. SR 4.2.3.2.c has boon added. The purpose of part c.1 is to perform margin calculations based on the measured radial l peak. The limit [ Fh (X , Y) )l* to which the measurement is l compared is based on the allowable design MARP limit, l
increased by a factor that represents the maximum amount that the power at the given assembly location can increase above the design value before the measured value may become limiting. Part c.2 uses the amount of margin dctormined by this proceduro to form the basis for the amount of power level reduction and the reduction in the high flux and OTAT K, trip sotpoints required in the ACTION statomonts for the specification. This is cc.nparable to - the now SR 4. 2. 2.2.c ' on F n(X,Y, Z) . i L 9. SR 4.2.3.2.d has boon added. This surveillance requires l projections of the measuromonts to be made to determino at
- What point Fu,(X,Y) would exceed the allowable limit if the current trend continues. In part d.1 a penalty is applied Ib-11 l.
{- 1 - --
to F$n(X,Y) if the trend indicates that the measured peak would exceed the limiting peak within the 31 EFPD surveillance period, and the margin calculations are repeated. This provides additional margin, or a buffer, to help ensure that the peak will not exceed the limit prior to next 31 EFPD measurement interval. In part d.2, the measurement is obtained and the margin calculations are repeated so that appropriate actions can be taken before zero margin is reached. This surveillance ensures the core is monitored at a frequency that considers conditions when measured peaks are underpredicted. This is comparable to the new SR 4. 2. 2. 2.d on F (X, n Y , Z) .
- 10. SR 4.2.3.3, 4.2.3.4, and 4.2.3.5 in the current specification address measurement of Reactor Coolant System flow rate. These requirements have been incorporated in Specification 3.2.5, and have been deleted from the revised requirements for SR 4.2.2.
_1schnical Justification Specification 3/4.2.3 was revised to provide required actions and surveillance requirements consistent with Duke Power company methodology for core power distribution control and surveillance of the nuclear enthalpy rise hot channel factor, as discussed DPC-NE-2011PA (Reference 2). The nuclear enthalpy rise hot channel f actor (Fm(X,Y)] is a specified acceptable fuel design limit that preserves the initial conditions for the most limiting non-0 TAT DNB transient (i.e., primary protection against DNB is not provided by the OTAT trip function). Fm (X , Y ) is defined as the ratio of the integral of linear power, along the rod with the highest integrated power, to the value of this integral along the average rod. Since Fm(X,Y) integrates the power along the length of the rod, it is related to the linear heat generation rate of the fuel rod, averaged over the length of the rod. When power distribution measurements from the incore detectors are obtained, the measured value of the assembly radial peaking f actor is labeled FTu(X,Y) . The Fm(X, Y) limits are preserved by the licensing design analysis, as described in DPC-NE-2011PA. Operation within the Fh(X,Y) limits defined in the COLR ensures that the measured peaking will be within the design calculations. The Fh(X,Y) limits are derived from the Maximum Allowable Radial Peaking (MARP) limits specified in the Core Operating Limits Report (COLR). The MARP limits are a family of maximum allowable radial peaking curves, typically plotted as maximum allowable radial peak versus axial location of peak, parameterized by the axial peaking factor. The family of curves is the locus of points for which the minimum DNBR is equivalent to that calculated for the most Ib-12 i
limiting non-OTAT transient (based on the reference design peaking). The MARPs in the COLR are based on the stateroint that represents the point of minimum DNBR during this transient. One set of MARP limits are specified in the COLR for Mark-BW fuel and OFA fuel. The DNBR penalty for optimized fuel in mixed cores is applied against the margin included in the design DNBR limit. The roload maneuvoring analysis determinos limits on global core paramotors that can be measured directly. The primbry parameters used to monitor and control the core power distribution are control bank insortion, axial flux differenco (AFD), and quadrant power tilt ratio. Limits are placed on those parameters to ensure the core power peaking factors romain bounded durir.g power operation. Uncertainties for the nuclear design model, engineering hot channel factor, assembly spacing, axia2 peaking factor, and other uncertainties in the CilF correlation were statistically combined to produce an overall DNDR uncertainty. This overall uncertainty was used to establish the statistical DNBR limit (SDL), as described in DPC-NE-2004 (Referenco 5). Since the MAP limits link the peaking limits to the DNBR limit, those uncertainties are accounted for in the MAP limits, and are not applied explicitly in the nanouvering analysis. Comparisons of the measured coro power distribution at steady-stato conditions to the design power distribution provido confirmation that the measured Fm(X,Y) is within the values of c~he designed core power distribution. This comparison verifica the applicability of the measured coro condition to the designed condition, so chat if the control bank insertion, AFD, and excore quadrant power tilt ratio are at tbsir most limiting values, then the-initial condition DNB peaking critoria are preserved. When the measurement is ootained, values of measured Fui(X,Y) are compared directly to the MARP limits for the LCO and compared-to (Fgn(X,Y))surv in the surveillance portion of the specification. (Fw(X,Y)]"" is obtained by adjusting the design F2n(X,Y) for the least amount of inargin obtained from all the power distributions analyzed in the maneuvering analysis. The survoillance methodolegi used ' in obtaining [ Fin (X,Y) ]"" is discussed in further detail in DPC-NE-2011PA (Reference 2). The resulting (fin (X,Y))"" and MARP limits are documented in the COLR. If FI,n ( X , Y) is less than (Fin (X,Y) )"", then positivo margin exists and Fm(X,Y) is within its limit. Pronosed Revision to Technical Specificati9n 3/4.2 4 This' revision applies to Unit 1 only. Specification 3.2.4 wat revised to reflect the requirement to i Ib-13 I _ . . . - - . . - . _ . - . - - _ . _ --- - - _ _ _ _ _ , - . .-_._ _ ~. ~____--_ _ _ ,,..-.-
decreano thermal power by at least 3% for each 1% of indicated quadrant power tilt ratio in excess of 1.02. This was donc because the power distribution analysis includes a peaking allowanco for quadrant power tilt ratios up to 1.02. When the quadrant power tilt ratio increases abovo 1.02, reductions in thermal power are required to limit the maximum local linear heat rato. The actions re'lluired to reduce thermal power are provided in the curront specification. Thorofore, this change in the specification reflects a quadrant power tilt ratio of 1.02 as the "referenco" value, abovo which a thermal power reduction is required. This revision is consistent with the peaking allowanco for quadrant power tilt, as duacribed above. . The Applicability of Specification 3.2.4 la for Modo 1 operation above 50% of rated thermal power. A statomont that "The provisions of 3.0.4 are not applicable" was also added to clarify that the surveillanco requirement would be completed above 50% of RATED.TilERMAL POWER. This chango is.administrativo in nature because it does not represent an actual chango to the requirements of Specification 3.2.4 or to its required actions. The revision to the Applicability Section applies to Unit 1 and Unit 2. A footnoto was added to the Applicability statomont to indicate that too specification is not applicable until completion of excoro detector calibration subsequent to refueling. Technical Justification Specification 3.2.4 was revised to provido required actions consist with Duke Power Company methodology for core power distribution control as discussed in DPC-NE-2011PA (Referenco 2). D'uring the reload licensing analysos for McGuiro Unit 1, Cycle 8, a three-dimensional maneuvering analysis was performed to datormine the core operating limits based on the methodology described in DPC-NE-2011PA. The analysis addressed the Axial Plux Differenco (AFD) limits (Specification 3.2.1), the control bank-insertion limits (Specification 3.1.3.6), and the Quadrant Power Tilt Ratio (Specification 3.2.4). The current. control bank insertion limits and AFD limits woro verified to be acceptable for the Cycle 8 core design. The control bank-insertion limits-and AFD-limits datormined-during the maneuvering analysis include an allowance for the quadrant power tilt-in the coro.- Specifically, the-calculated peaking is' increased by an amount corresponding to a 2% quadrant power tilt, equiva!'nt to a quadrant power tilt ratio of 1.02. Thorofore, the ror;1 ting control bank insertion and AFD limits are valid for excara quadrant power tilt ratios up to the Technical Specification value of 1.02. Simulations of quadrant-Ib-14
i power tilts and corresponding peaking increases have shown that the peaking allowance described in DPC-NE-2011PA is sufficient to bound the increased peaking due to tilt up to a quadrant power tilt ratio of 1.02. This peaking allowance has been used in the maneuvering analysis to determine both AFD (operational) limits and f(61) (safety) limits. EIOPJ2fLtil_RDvision to Technical Soccification 3/4.2.5 This change applies to Unit 1 only. This proposed Technical Specification (TS) revision changes TS 3.2.5 to include a now figure that defines power reduction requirements for low flow operation. This now figure, Figure 3.2-1 replaces Figure 3.2-3 which had previously been relocated from Specification 3.2.3 to Figure 8 of the COLR. Technical Justification Since maximum allowable peaking limits are still defined in Specification 3.2.3, the peaking dependence that was previously included on COLR Figure 8 in the parameter R has been climinated from the new figure. Specification 3.2.5 has also been revised to include the action statements from Specification 3.2.3 that govern the response to the power and flow combination being in the regions of restricted or prohibited operation, while maintaining the current action statement that governs the response to temperature and pressure exceeding their limits. The region of acceptable operation was redefined as regions of permissible and restricted operation in the new specification to better describe allowed regions of operation. Action b. and c.1.b were incorporated in Specification 3/4.2.5 to maintain proportionality betwoon %RTP and the Power Range Neutron Flux-High Trip Setpoint. In addition, all flow rate surveillance requirements previously contained in Specification 4.2.3 have been moved to Specification 4.2.5. Although Figure 3.2-3 had previously been moved to the Core Operating Limits Report (COLR), the new figure, Figure 3.2-1, is being returned to the plant Technical Specifications, since the parameters that are governed by the figure (power and flow) do not change on a cycle by cycle basis. Previously, Figure 3.2-3 in Specification 3.2.3 related minimum flow as a function of both power and R, where R was function of the Nuclear Enthalpy Rise Hot Channel factor, power, and two cycle specific parameters (the design Nuclear Enthalpy Rise Hot Channel factor and the low power peaking adjustment factor). Because Figure 3.2-3 contained cycle specific parameters, it was moved from Specification 3.2.3 to the COLR. The now figure, Figure 3.2-1, relates minimum flow only to power. The dependence Ib-15
of minimum flow on R has boon eliminated. Thorofore, since the dependence of the figure on cyclo specific paramoters no longer
- oWists, it can be returned to the Technical Specifications.
Figuro 3.2-1 defines the trado-off in power and flow that will maintain the bases of the coro safety and operating limits for a low flow condition. The analysis that verifies this trado-off considorod a flow reduction down to 90 percent of the thermal design flow rato. To verify the validity of the trade-off, ' savoral DNBR ovaluations woro performed. Those ovaluations demonstrate that the overtemperaturo AT safety limits remain valid as flow is reduced and that the specified reduction in oporating power loval produces improved DNBR margins for the limiting condition II ovent. All reduced power statopoints considered the increase in allowable core peaking at reduced , power consistent with Technical Specification 3.2.3. Thorofore, no additional limits on the maximum Nuclear Enthalpy Rise Hot Channel Factor are required duo to reduced flow conditions, thoroby removing the R dependence. To ensure that the lovel of protection that has boon assumed in the plant safety analysis licensing basis safety analyses is maintained, all action statomonto that woro previously included in Specification 3.2.3 have boon retained in the now Specification 3.2.5. This assures that non-loss-of-flow transients, like the rod withdrawal at power, are protected at the low flow condition. Erpposed Raylnign to TechnicALJngqiLLqat.ipn_TAhips 3. 3-1dd-2
& 4.3-1 This chango applion to Unit 1 only.
This change is to doloto t80 reactor trip on Power Range Neutron Flux.Nogative Rated from the Reactor Protection System. Toehnica1 Jyntifinalign Refer to the technical justification for the proposed revision to TS Tablo 2.2-1. Proposed Revision to TechDigal Specification Table 3.3-4 This change applies to Unit 1 only. This proposed revision changes the low steam lino prosauro sotpoint for safety injection and main steam lino isolation from 585 psig to 775 psig. Also, the allowable value for this trip function is changed from 5,65 psig to 755 psig, maintaining the namo 20 psig allowance for rack uncertainties. Technical Justificat.Lon Ib-16
{ The higher steam lino pressure sotpoint is consistent with all licensing basis safety analyses. The steam lino break event was reanalyzed (Reference 4) assuming an uncompensated low steam line pressuro sotpoint of 700 psig, allowing 73 psig for instrument uncertainty and margin. This reanalysis shows this Condition IV transient does not exceed the imposed Conditic, II acceptanco critorion of no DNB. The inadvertent opening of a steam generator relief or safety valvo, in terms of primary system overcooling, is essentially a small steam line break. This Condition II event exhibits steam releases markedly loss than the steam lino break event. Thorofore, this event is bounded by the steam line break event and does not requiro reanalysis. This chango does not necessitato reanalysis of the peak containment temperature analysis. The removal of the dynamic compensation of the steam lino pressure signal, even with the increased setpoint, will delay main steam lino isolation following the break in the main steam line. Since the offect of main steam isolation is to terminato the blowdown of the intact steam generators, postponing MSIV closure will tend to rotard the blowdown of the faulted generator and delay tubo uncovery. Also, tho extended blowdown of the intact steam generators will reduce the Reactor Coolant System temperature, which will further delay tube uncovery. Once uncovery does occur, this lower primary system temperature will offectively reduno the enthalpy of the break flow. The removal of thm Tic compensation of the steam pressure signal, which accomp, .a the change in the low pressure notpoint, will ossentA 11y eliminate the spurious ESP actuation on minor (but rapid) r ensure decreanos in the secondary system. An interim modification, which reduces the lead-lag timo constants from 50-5 to 2-1 rather than removing all dynamic compensation, will not invalidato any of the above conclusions. Steam line isolation will be delayed due to the largo reduction in the " lead" timo constant. PIgposed Revision tp Technical Specification _Ipble 3.3-5 This change applies to Unit 1 only. 1No responso timos are modified in this proposed change, the feedwater isolation responso time is changed from 9 seconds to 12 seconds and the steam lino isolation time is changed from 7 noconds to 10 seconds. Ig_qhnical JustificatiQD The extended response times are consistent with or conservative Ib-17
for all licensing basis safety analyses. These two response timos are assumptions in the steam lino break analysis. Increasing those responso timos causes the primary system overcooling to worson duo to the extended blowdown of the intact generators and the addir.ional mass of main foodwater delivered to the faulted generator. Roanalysis (Reference 4) shows this Condition IV transient doon not exceed the imposed Condition II acceptance critorion of no DNB. The inadvertent opening of a steam generator relief or safety a valvo, in terms of primary system overcooling, is essentially a I small steam line break. This Condition II event exhibits steam releases markedly loss than the steam line break event. Thorofore, this event is bounded by the steam line break event and does not require roanalysis. The, increased foodwater isolation response timo also impacts the analysis of the excessive foodwator flow event. The offect is negligible, however, since the DNBR decreaso is terminated by the Lturbino trip, which occurs 3 seconds after the high-high cteam generator level setpoint is reached. The fact that the foodwater-isolation _ valves close 12 seconds after this setpoint is reached rather than 9 seconda does not affect the minimum DNBR achieved. For the peak containment temperature analysis IReference 6), a 4 lower liquid mass in the faulted generator yields conservativo results. This loads _to an earlier uncovery of the steam generator tubes and, thus, the advent of superheated steam oxiting the break. Longthening the foodwater isolation response time will increase the amount of foodwater delivered to the faulted generator and delay tubo uncovery. Increasing the main steam line isolation response time has a similar offect as the removal of the dynamic componsation in the steam line pressure signal discussed in tho *.ochnical justification for the proposed change to TS Table 3.3-4. Significant difficulty has boon experienced in mooting the current specification response times for both of those ESP functions. Increasing the allowable response times should climinate this difficulty. Eroposed Revision to Technical Specification 3.4.1.2
-This change applies to Units 1 and 2.
-Tho= specification is being changed to require that the throo operable reactor coolant loops be in operation in Mode 3.
Technical Justification This restriction is imposed in order to make the specifications Ib-18
consistent with the analysis of the uncontrolled bank withdrawal from subcritical or low power startup condition. Propog_ed Revision to Technical Specification lei.2.1 & 3.4 tl12 This chango applies to Unit 1 only. , This modification changes the tolerancos on the pressurizer safety valvo lift sotpoint from ilt to +3%,-2% in all modos of operation. Igghnical Justification The pressurizar code safeties are not tested in place but are removed and shipped to a testing facility. The safety concerns for removal and replacement of those valvos are difficult access to the work area, difficulty in lifting device rigging for valve removal / replacement, and valve trarsport to/from the pressurizer. Since this work is performed in a radiological environment, work activitios are further complicated by anti-contamination clothing. For a conservativo approach, all three valvos are removed each outage for testing. The setpoint drift soon during testing would again fall under '.no proposed sotpoint varianco change. The change would possibly reduce work in the pressurizor by 66% by requiring only one valvo to be tested por outage. In summary, safety bonofits would be gained by loss work in a dangerous environment and loss radiation exposure. The larger allowable deviation from the nominal lift setting is consistent with the licensing basis analyses. An increased proasurizer safety valvo lift setpoint impacts the peak Reactor Coolant System pressure calculated for pressure increase transients. A pressure increase is the result of a heatup in the Reactor Coolant System due a mismatch between the heat generated in the reactor core and the heat removed by the secondary system. The three accident categories involving such heat transfer mismatches are the decrease in secondary heat removal, decrease in Reactor Coolant System flow rate, and reactivity and power , distribution anomaly transients. The feodline break, locked ; rotor and rod ejection events are the limiting pressure increase transients in these three accident categories, respectively. These events have all been analyzed assuming a lift sotpoint 3 I -percent above the nominal value. These analyses show that the peak Reactor Coolant System pressure criterion (310% of design pressure) is met for each event. L The amount by which the safety valve lift setpoint is allowed to drift downward is restricted to 2 percent of nominal in order to ensure that safety valvo lift cannot preclude reactor trip on high pressurizer pressure. For DNB transients in which a high pressurizer pressure reactor trip does not provent the lifting of l Ib-19 _ ___ . _ _ . . _ _ _ - _ . . ~- -
l the safoty valves, the offect of this reduced setpoint on the transient DNBR is ovaluated. Sinco low pressuro is conservative for DNBR analyses, it is typically assumed that the pressurizer PORVs and sprays mitigato the pressure increase due to the system hoatup and thereby preclude safety valvo lift. For the uncontrolled bank withdrawal at power and single rod withdrawal events, however, the operation of the pressurizer pressure control system would tend to ylold an earlier reactor trip on overtomparaturo AT due to pressuro compensation of the trip sotpoint. The roanalysis of thoso events show that all acceptance critoria are met.
)
Eroposed Revision to Technical Snecification 3.5.1.1 This change applies to Units 1 and 2. This change raises the required averago cold log accumulator
- boron concentration in ACTIONS c.2 and c.3 from 1500 to 1800 ppm, and bases this average on all four accumulators instead of just the limiting throo. Also, in each of the ACTIONS, "pressurizor" is administratively changed to " Reactor Coolant System."
Isshnical Justification Calculating the volumetric average boron concentration based on all four cold leg accumulators is valid, since, regardless of the break location, the contents of each accumulator will be emptied , (either directly or indirectly) into the containment sump. A volumotric average concentration of 1800 ppm will ensure long-term subcriticality following a LOCA. Changing " pressurizer" to " Reactor Coolant System" is administrativo in' nature. The change is made to reflect the instrument actually used by the plant to complete the required ACTIONS. Pressurizer pressure goes off scale low at 1700 psig, so it cannot be used to measure pressures below 1000 psig as stated in the current TS. Pronosed Revision to Technical Specification Table 3.7-3 This change applies to Unit 1 only. Similar to the proposed changes to specifications 3.4.2.1 & 3.4.2.2, the tolerance on the steam line safety valvo lift setpoints is being raised from 11% to i3%. TDShnica1 Justification The main steam code safety valves must be tested with full steam header pressure. The testing requires removal of the manual Ib-20
~ _ , _ _ . _ - _ , _ _ _ _ _
l actuation device and installation of an air motor assembly for each valve to be tested. All of these components are installed in top of the doghouses. The test code as written requires a sample of these valves to be tested at each shutdown and if test failure occurs, then the population must be expanded. If other failures occur, the test population must be further expanded. The worst setpoint drift has been approximately 10 psi. This is a failure under the present 1% criterion, but would easily satisfy 3%. In summary, the safety benefit from this change would be fewer manhours spent for testing and rework in a dangerous environment (high temperature / pressure piping, high elevation, difficult access). The larger allowable deviation from the nominal lift setting is consistent with the licensing basis aralyses. An increased safety valvo lift setpoint directly impacts the Main Steam System peak pressure transients. The turbine trip, MSlV closure and uncontrolled bank withdrawal events are analyzed in order to ensure that the 110% of design pressure acceptance criterion is not exceeded. EI9PMfi BEi.nlon to Ttchnical Spfgjfication 4.7.lM This change applies to Unit 1 only. The permissible stroke time for the main steam isolation valves is being changed from 5 to 8 seconds. Technical JustificatioD The larger allowable isolation valve stroke time is consistent with or conservative for all licensing basis safety analyses. The valve stroke time, when added to the applicable instrumentation delays, yields the overall ESF response time. As stated in the technical justification for the proposed revision to TS Table 3.3-4, this response time is input to the steam line break transient analysis. Increasing the stroke time causes the primary system overcooling to worsen due to the extended blowdown ' of the intact generators. Roanalysis (Reference 4) shows this Condition IV transient does not violate the imposed Condition II acceptance criterion of no Dt1B. The inadvertent opening of a steam generator relief or safety valve, in terms of primary system overcooling, is essentially a small steam line break. This Condition II event exhibits steam releases markedly less than the steam line break event. Therefore, this event is bounded by the steam line break event and does not require reanalysis. For the peak containment temperature analysis (Reference 6), increasing the main steam line isolation response time has a i similar effect as the removal of the dynamic compensation in the l I Ib-21 l
steam lino pressure signal discussed in the technical ) justification for the proposed change to TS Table 3.3-4. Significant difficulty has boon experienced in mooting the current specification stroke timo for those valves. Increasing the allowable stroke timo should eliminate this difficulty. Ergng_ pod Revision to Technical SpecifigatJon 4.5.2 ( f &_Ill This chango applies to Units 1 and 2. T. S. 4.5.2(f) gives the ECCS pump performance requirements. The contrifugal charging pump required developed head is decreased from 2380 to 2339 paid. The safety injection pump required developed head is increased from 1430 to 1454 paid. The residual ) heat removal pump required developed head is increased from 160 , to 169 psid. T. S. 4.5.2(h) gives the ECCS delivered flow requirements. The contrifugal charging pump delivered flow under I LOCA conditions at atmosphoric pressure is decreased from 345 to 335 gpm.- The cafety injection pump required flow is decreased : from 462 to 405 gpm. Those revisions are applicable to Unit 1 and Unit 2. Technical Justification The ECCS pump required developed head and delivered flow specifications-are being revised to provido performanco test margin for periodic testing, and to be consistent with the current safety analysos and with the rovised analysos associated with McGuire 1 Cyclo 8. performance test margin is required to onable sufficient allowances for instrument errors and to permit reasonable test acceptance critoria, pump performance at the now specification values is sufficient to moot all acceptanco critoria in both the current FSAR analysos, and in the revised analysos submitted with or referenced by the McGuiro 1 Cycle 8 roload. The proposed revisions are therefore acceptable. Ib-22 _ . _ . _ . . ._._-... _ _ _ _ . _ u.____.._. - - - _ _ _ _ _ _ . . -__ _ _ _ _ _ _ . - _ . _ _ _
B(liflIRI19211
- 1. BAW-10159P-A, BPCMV Correlation of Critical licat Flux in Mixing Vano Grid Fuel Assemblics, Babcock & Wilcox, July 1990.
- 2. DPC-NE-2011PA, Duke Power Company, Nuclear Design Methodology for Coro Operating Limits of Westinghouse Reactors, March, 1990.
- 3. BAW-10174A, Mark-BW Roload LOCA Analysis for the Catawba and McGuire Units, Babcock & Wilcox, Revision 1, Approved May, 1991
- 4. DPC-NE-3001P, Duko Power Company, Multidimensional Roactor Transients and Safoty Analysis Physics Parameters Methodology, January 1990.
- 5. DPC-NC-2004, Duke Power Company, McGuire and Catawba Nuclear Stations Core Thermal-llydraulic Methodology using VIPRE-01, December 1988.
- 6. WCAP-10988, Cobra-NC, Analysis for a Main Steamlino Break in the Catawba Unit 1 Ice Condonsor containment, Westinghouse Nuclear Energy Systems, November 1985.
Ib-23
Attachment Ic No Significant llazardo Analyses Q9NTENTS: POWER DISTRIBUTION AND SAFETY LIMITS (PAGE 2) REMOVAL OF NEUTRON FLUX !!IGli 11EGATIVE RATE (PAGE 6) INCREASE IN REQUIRED NUMBER OF LOOPS Ill OPERATION (PAGE 7) COLD LEG ACCUMULATOR REQUIRED BORON C011 CENTRATION INCREASE (PAGE 7) Cl!ANGES TO ECCS PUMP PERFORMANCE REQUIREMENTS (PAGE 8) INCREASE PRESSURIZER AND MAIN STEAM CODE SAFETY VALVE SETPOINT TOLERANCES (PAGE 9) LOW STEAMLINE PRESSURE SETPOINT CilANGE (PAGE 10) FEEDWATER AND MAIN STEAM ISOLATION RESPONSE TIMES (PAGE 11) ADMINISTRATIVE CilANGES TO CilAPTER 6 (PAGE 12) CilANGE TO LIST OF ACCIDENTS W111Cll MUST BE REEVALUATED IN Tile EVENT OF AN INOPERABLE RCCA (PAGE 13) ENVIRONMENT IMPACT STATEMENT (PAGE 14)
110 SIGNIFICANT II/GARDS ANALYSIS SAFETY LIMITS AND POliER DIETRIBUTION TECilillCAL SPECHLCATlDim The following analysis, required by 10CFR 50.91, concludes that the proposed amendment will not involvo significant hazards considerations as defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increano in the probability or I consequences of an accident previously evaluated; or
- 2) Create the possibility of a new or different kind of accident from any previously evalcated; or
- 3) Involve a significant reduction in the margin of safety.
The fuel for McGuiro Nuclear Station Unit 1 Cycles 1-7 was supplied by Westinghouse. As a result of Duke Powor's decision to open future reload contracts to competitive bidding, the fuel for at least Cycles 8-10 of McGuire Unit I will be supplied by B&W Fuel Company. Catawba Unit 1 Cycle 6 was the first Duke Power nuclear station for which BWFC supplied the roload fuel. The Catawba Unit 1, Cycle 6 Reload Report (Reference 1) presented an evaluation which demonstrated that the core roload using Mark-BW fuel will not adversely impact the safety of the plant. This reload report for McGuire is similar to Reference 1, but reflects that the analyses performed in support of the operation of Cycle 8 have been performed by Duke Power, rather than BWFC. During Cycle 8 the core will contain 76 fresh fuel assemblies supplied by B&W and 121 Westinghouse supplied Optimized Fuel Assemblies (OFA). Methods and models have been developed to support McGuire Unit 1 Cycle 8 operation during both normal and of f normal operation. These methods and models ensure safe operation with an entire core of Mark-BW fuel and with a core of mixed Westinghouse and Mark-BW fuel. The analysis methods are documented in Topical Reports which have boon submitted to the NRC, and are either under review or approved. These Topical Reports are listed in Section 10 of Attachment 3. For the reload-related Technical Specifications the probability or consequences of an accident previously evaluated is not significantly increased. A LOCA ovaluation for operation of McGuire Nuclear Station with Mark-BW. fuel has been completed (BAW 10174, Mark-BW Roload LOCA Analysis for the Catawba and McGuiro Units). Operation of the station while in transition from Westinghouse supplied OFA fuel to Ic-2 i
B&W supplied Mark-BW fuel is also justified in this topical. 4 BAW 10174 demonstrates that McGuire Nuclear Station continues to meet the criteria of 10 CPR 50.46 when operated with Mark-BW fuel. Large Break LOCA calculations completed consistent with an approved evaluation model (BAW 10168P and revisions) demonstrate compliance with 10 CFR 50.46 for breaks up to and including the double ended i severance of the largest primary coolant pipe. The small break ^ LOCA calculations used to license the plant during previous fuel cycles are shown to be bounding with respect to the now fuel design. This demonstrates that the plant meets 10 CPR 50.46 criteria when the core is loaded with Mark-BW fuel. . During the transition from Westinghouse OPA fuel to Mark-BW fuel both types of fuel assemblics will reside in the core for several fuel cycles. Appendix A to BAW-10174 demonstrates that results presented above apply to the Mark-BW fuel in the transition core, and that insertion of the Mark-BW fuel will not have an adverse , impact on the cooling of the Westinghouse fuel assemblies. Duke Power Company's Topical Reports DPC-NE-3000, DPC-NE-3001, and DPC-NE-2004 provide evaluations and analyses for non-LOCA transients which are applicable to McGuire. The scope of those analyses includes all events specified by sections 15.1-15.6 of Regulatory Guide 1.70 (Standard Format and Content of Safety Analysis Reports for Nuclear Power Planta) and presented in the Final Safety Analysis Report for McGuire. The analysis and evaluations performed for these topicals confirm that operation of McGuire Nuclear Station _for reload cycles with Mark-BW fuel will continue to be within the previously reviewed and licensed safety limits. One of the primary objectives of the Mark-BW replacement fuel is compatibility with the resident Westinghouse fuel assemblics._ The description of the Mark-BW fuel design, and the thermal-hydraulics and core physics performance evaluation demonstrate the similarity between the reload fuel and the resident fuel. The extensive testing and analysis _ summarized in BAW 10173P (see Reference- 1) shows that the Mark-BW fuel _ design performs, from the standpoint of neutronics and thermal-hydraulics, within the bounds and limiting design criteria applied to resident Westinghouse fuel for the Catawba plant safety analysis. Each FSAR accident has been evaluated to determine the effects of Cycle 8 operation and to ensure that the radiological consequences
.of hypothetical accidents are within ~ applicable regulatory guidelines, and do not adversely affect the health and safety of
-the public. The design basis LOCA evaluations assessed the radiological impact of differences between the Mark-BW fuel and Westinghouse CFA fuel fission product core inventories. Also, the dose calcubtion effects from non-LOCA transients reanalyzed by BWFC utilizing Cycle 8 characteristics were evaluated. Differences Ic-3
1 in the current FSAR dose values that are not related to the insertion of Mark-BW fuel reflect the application of the latest revisions to Standard Review Plan dopo assessment methodology. The calculated radiological consequences are all within specified regulatory guidelines and contain significant levels of margin. The analyses contained in the referenced Topical Reports indicate that the existing design criteria will continue to be met. Therefore, these TS changes will not increase the probability or consequences of an accident previously evaluated. As stated in the above discussion, normal operational conditions and all fuel-related transients have been evaluated for the use of Mark-BW fuel at Catawba Nuclear Station. Testing and analysis was also completed to ensure that from the standpoint of neutronics and thermal-hydraulics the Mark"BW fuel would perform within the limiting design criteria. Because the Mark-BW fuel performs within the previously licensed safety limits, the possibility of a new or different accident from any previously evaluated is not created. The safety analyses performed in support of any reload necessarily l involve the assumption of a number of input parameter values. Because of the differences in methodologies used by the various analysts, and the proprietary nature of the analyses, a side-by-side comparison of input assumptions is generally neither possible nor useful. The reload-related changes to the TS do not involve a significant reduction in the margin of safety. The calculations and evaluations documented in BAW 10174 show that McGuire will continue to meet the criteria of 10 CFR 50.46 when operated with Mark-BW fuel. The evaluation of non-LOCA transients documented in DPC-NE-3001 also confirms that McGuire will continue to operate within previously reviewed and licensed safety limits. Because of this, the TS changes to support the use of Mark-BW fuel will not involve a significant reduction in the margin of safety. The technical changes made to Table 2.2-1 reflect the use of the BWCMV CHF correlation and Duke Power's Statistical Core Design methodology with a 1.55 thermal design limit. These changes to Table 2.2-1 will not significantly increase the probability or consequences of an accident previously evaluated. The changes to the K values conservatively bound the allowable operating region, as defined by the new DNBR methodology. It can be concluded that these changes will not create the possibility of any new accident from those previously evaluated. It can also be concluded that since all new TS values are bounded by safety analysis assumptions that this change vill not significantly decrease the margin of safety. Several of the requested amendments are administrative in nature. The requested change which updates Table 2.2-1 for deletion of the Ic-4
i RTD Bypass System, reflects a change which has been previously approved by the NRC ( Amendment No. 84 to Facility Operating License NPF-9 and Amendment No. 65 to Facility Operating License NPF-17) . Since the needed modifications have boon completed on both McGuire units the reference to the manifolds is obsolete and is being doloted. Since there is no change in requirements this change does not involve significant hazards considerations. An administrative change is being made to the TS which identify which TSs apply to Unit 2, and no longer apply to Unit 1 after the reload. Table 2.2-1 has been laboled to reflect the unit to which it applies. The power Distribution TS (3/4.2.1, 3/4.2.2, 3/4.2.3, 3/4.2.4) have been similarly labeled to specify unit-specific applicability. The existing TS will be copied on yellow paper to further distinguish them from the new TS which apply to Unit 1 only. The Power Distribution TS will have an "A" in the page number for Unit 1 and a "B" for Unit 2. The pages will also be marked " Unit 1" or " Unit 2". This change is administrative only, and is being made to distinguish between the TS for Unit 1, which will be operated with TS revisions which reficct the use of Mark-BW fuel, and Unit 2 which will continue to operate with Westinghoose supplied fuel. Based on the above, it is concluded that no significant hazard considerations exist. REEEREllCE1 Catawba Nuclear Station Technical Specification Amendment Application, M. S. Tuckman to USNRC, January 9, 1991. Ic-5
.___m NO SMidf1 CANT IIMARDS_1WAIJEIS DILETION OF NEU_ TRON liMll_J1EGATIVE IMTE TIUP iTSJ ABLES 2.2-1. 3.3-1, 3.3-2 JiRD_4.3-1)
The following analysis, required by 10CFR 50.91, concludon that the proposed amendment will not involvo significant hazards considorations as defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2) Croato the possibility of a now or different kind of accident from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safety.
The removal of the Power Range Neutron Flux liigh Negativo Rato trip will not result in any previously-reviewed accident becoming more probable or more sovero. The trip is a responso to a pre-existing transient condition and would not initiato any accident. The trip is designed to provido protection from a dropped control rod. However, in the ovent of a dropped rod the reactor is assumed to trip, if a trip is to occur, on low pressurizer prosauro. Thorofore the protection function is retained. The consequences of a dropped rod have been analyzed and found to be within acceptablo limits. Likewise, the removal of this trip will not create a now accident not previously reviewed. The removal of a response to a transient will not initiato a now transient. There are no crediblo unanalyzed transients which will occur as a result of a dropped rod. The removal of this trip will reduce the potential for spurious or unnecessary trips which may occur as a result of maintenanco or the drop of a low-worth rod. There are no other hardware modifications or proceduro changes whicn are to be made at. a result of the deletion of this trip function which could create the possibility of a now accident. No margin of safety ' will be reduced by this chango. As noted above, if a dropped rod necessitates a trip, the trip function will be accomplished as a result of low pressurizor pressure. For those dropped rods for which no trip is necessary, the removal of this trip will provido protection against an unnecessary transient. Based on the above, it is concluded that no significant hazard considerations exist. Ic-6 i i I
NO SLONIFICANT !!AZARDS ANALYSIS __lHCREASE NUMBER OF OPERABLE RCS LOOPS.IN MODE 3 (TS 3.4.3.2)_ and JNCREASE COLD LEG ACCUMULATOR REOUIRED BORON CONCENTRATION (TS 3.5.1.1) The following analysis, required by 10CFR 50.91, concludes that the proposed amendments will not involve significant hazards considerations as defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations (NSilC) if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2) Create the possibility of a new or different kind of accident from any previously evaluated; or 3)- Involve a-significant reduction in the margin of safety.
Those amendments will not involvo any significant hazards consideration. The proposed changes will result in the parameter or operating condition involved to become more restrictive (conservative) than - currently exists. The NRC's own guidanco, published in the Federal Register (48CFR 14870) states that an amendment which results conditions becoming more restrictive-are not likely to result in an NSitC. Therefore, it may be concluded with no further analysis that those amendments will not involvo a Significant Hazards Consideration. Ic-7
up_IiLG111FICAt4T 11AZAnas_AHALyplfi ECCS_PJ}iP P1:RF.ORMA11CE IWQElRiliEti l li (TS_4.5.2 f Lh1 The f ollowing analysis, required by 10CFR 50.91, concludes that the proposed amendments will not involve significant hazards consid* rations as defined by 10 CFR S0.92. 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations (11SilC) if operation in accordance with the propoced amendment would nott
- 1) Involvo a significant increase in the probability or consequences of an accident previously evaluated; or
- 2) Croato the possibility of a now or different kind of accident from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safety.
The proposed amendments will not involve a significant increase in the probability or consequences of an accident previously evaluated because the Loss-of-Coolant-Accident (LOCA) analysis, to which the flowrates aro input assumptions, continue to meet applicable acceptance critoria. The proposed amendments will not result in a significant increase in the possibility of a new accident because the now values represent a change in assumptions made in the LoCA analysis, rather than a physical chango in the plant. The proposed changes will not result in a significant decrease in a margin of safety, because pump performance at the new values is suf ficient to meet all acceptanco criteria in both the current FSAR analyses and in the revised McGuire 1 Cycle 8 analyses. Based on the above, it is concluded that no significant hazard considerations exist. Ic-8 I
liq _S I G N I PLQMiT ll A Z ARDS_M!AliijilS 11K11 EASE PRESEURlZEILAHILliA11LSTEhtLCDDILShi'liTY VAIRE SETPOINT TOLElmlNES ITS 3.4.2.1__&__2,Tablo 3.7-31 Tho following analynin, required by 10CFR 50.91, concludes that the proposed amendments will not involvo algnificant hazardo considerations an defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no significant hazards conaldorations (NSilC) if operation in accordance with the proposed amendment would not:
- 1) Involvo a significant increase in the probability or consequencon of an accident previously evaluated; or
- 2) Create the possibility of a now or different kind of accident from any previously ovaluated; or
- 3) Involvo a_nignificant reduction in the margin of safety.
The proposed amendment will noF. result in a significant increano in the probability or consequences of any previously analyzed accident. The valvo lift sotting is challenged only after a transient has boon i nitiated and is not a contributor to the probability of any transient or accident. The tranolonto which involvo pressure increanon which would pote7tially challengo the uafoty valvoo have boon analyzed to determino the conooquences of dolayed or prematuro valvo actuation at the extremen of the now actpoint tolerances. These analycos show that all applicable acceptance critoria are mot using the wider tolerances. The proposed amendment will not result in the creation of any now accident not previously evaluated. As noted above, the actpoint tolerance only affects the timo at which the safety valvo opens following or during a transient, and is not a contributor to the probability of an accident. Tho proposed amendment will not result in a significant decreano in a margin of safety. The limiting transient in each accident category has boon analyzed to determine the of rect of the chango in lift sotpoin; tolerance on the transient. In each caso, the results of the analyson mot all acceptanco critoria. Based on-the above, it is concluded that no significant hazard considerations exist. Ic-9
,we w--.. -m--a, r,-,,e a ,- -,--m,- -,-,--m --
110 SIGli1PICAllT llAZARDS AliAMf9JR LOW STEAM LIJ{E_fJlEEfiMRE SETPolliT CilAliGE (TS TABLE 3J -4) The following analysis, required by 10CFR 50.91, concludos that the proposed amendment will not involve significant hazards considerations as defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no l significant hazards considorations if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2)Croato the possibility of a new or different kind of accident from any previously evaluated; or 3)lnvolvo a significant reduction in the margin of safety.
Changing the Low Stoam Lino Prosauro sotpoint will not increase the probability or consequences of any previously-reviewed accident. The higher steam lino pressure notpcint is consistent with all licensing basis safety analysos. This change, in conjunction with the removal of the dymanic componsation of the steam pressure , signal, is intended to reduce or eliminato spurious Engincored Safeguards Features (ESP) actuations which are caused by minor (but rapid) pressure decreases in the secondary system. The proposed amendment will not result ir a now accident not previously reviewed. A change in steam lino peosauro is a response to an existing transient condition, rather than a procursor or , initiating event. A chango of steam lino pressure sotpoint is also not a procursor or initiating event. Tno proposed amendment will not result in a significant decrease in a margin of safety. The roanalysis of the steam lino break accident which was performed shows that all imposed Condition II acceptance critoria mot. mot. Based on the above, it is concluded that no significant hazard considerations exist. Ic-10
)
l NO SIGN 1f1 CANT HAZARDS ANALYSI EEEDWATER AND MAlli. STEh!LlHOLATION RESPONSE TIM 113
.{TS TABLE 3.3-5)
A11D MAIN STEAM ISQLAT1911 VALVE STROKE TlliE
.LTS 4 . 7 .1. 4 )
The following analysis, required by 10CFR 50.91, concludes that the proposed amendment will not involvo significant hazards considerations as defined by 10 CFR 50.92. l 10 CFR 50.92 states that 3 proposed amendment involves no l significant hazards considerations if operation in accordance with the proposed amendmont would nott
- 1) Involve a significant increase in the probability or consequences
! of an accident previously evaluated; or 2)Croato the possibility of a now or dif ferent kind of accident , from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safoty.
t The pro posed changes will not significantly increase the probabil:.ty or consequences of any previously evaluated accident. The effects of the delays in isolation timos on the various transients affected have boon analyzed and found to be acceptable. The pro posed changes will not sigisift:antly increase the l possibil:,ty of a now accident not previous 1* ovalua.ted. Foodwater and main steam isolation are r;sponses 'o cogoing transients, rather than initiators or procursors of treasients. No equipment or component reconfiguration will oc;m r.,a result of this change. The proposed changes will not slynificantly decrease any margin of safety. As noted above, the oflects of the longer isolation times have boon ovaluated and found to be acceptable. l Based on the above, it is concluded that no significant hazard l considerations exist. l l 1 Ic-11 l l
NO SIGNIFICANT HAZARDS C011 SIDERATION ANALYSIji ADMINISTRATIVE CHANGES TO CilAPTER 6 (TS 6.9.1.9) The following analysis, required by 10CFR 50.91, concludes that the proposed amendment will not involve significant hazards considerations as defined by 10 CFR 50.92. 60 CFR 50.92 states that a proposed amendment involves no oignificant hazards considerations if operation in accordance with the proposed amendment would not* 1
- 1) Involve a significant increase in the probability or consequences !
of an accident previously evaluated; or ! I
- 2) Create the possibility of a new or different kind of accident I from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safety.
These proposed changes to Technical Specifications are administrative in nature and as such will not involve a significant hazards consideration. The changes reflect the application of previously-approved Core Operating Limits Report (COLR) methodology for changing cycle-specific variables. COLR methodology was approved by the NRC for McGuire as Facility Operating License amendment nos. 105 (Unit 1) and 87 (Unit 2). Based on the above, it is concluded that no significant hazard considerations exist. I I Ic-12 i i
._ .- _ . . . _ . . . _ . ~ . . _ _ - . . _ _ . . _ -_ ..,. _ . _ . . . - _ _ . ~ . . _ _ . . - - .
NO SIGNIFIMIT llAZARDS CQL{SIDIIET1011 ANALYSIS REVISE LIST OF ACCIDENTS REOUIRING REEVALUATIOff IN THE FVEllT OF AN INOPERABLE RCCA (TS Table 3.1-1) The following analysis, required by 10CFR 50.91, concludes that the proposed amendment will not involve significant hazards considerations as defined by 10 CFR 50.92. 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2) Create the possibility of a new or different kind of accident from any previously evaluated; or
- 3) Involve a significant reduction in the margin of safety.
The proposed change to Table 3.3-1 will not change the probability or consequences of any accident or reduce any safety margin, because the table simply lists accident analyses which must be reevaluated in the event of an inoperable rod cluster control assembly (RCCA). The activities involved are analytical only, and do not introduce any operational considerations. Revision of the table to more accurately define the affected analyses is an administrative affort related to activities (analyses) which are conducted offsite after the fact of a postulated inoperable RCCA. Based on the above, it is concluded that no significant hazard considerations exist. Ic-13 I. _ __.___-_..m___
i Et1VIRON!EllTAL IMPACT STATEME![T The proposed TS changes have boon reviewed against the criteria of 10 CFR 51.22(c)(9) for environmental considerations. As shown , abovo, the proposed changes do not involve any significant hazards ! consideration, nor significantly incroaco the types or amounts of of fluents that may be released of fsito, nor significantly increase the individual or cumulative occupational radiation exposure. Based on this, the proposed Technical Specification chango moots the critoria given in 10 CFR 51.22(c)(9) for cotogorical exclusion from the requirement for an Environmental Impact Statomont. b f i Ic-14
Attachment II Changes to the Core Operating Limits Report
~ - _ - - - _ _ _ _ _ - _ _ _ _ _ _ _ _ . __ _ _ _ _ _ _ _ _
N .I.h b McGuire 1 Cycle 8 Core Operating Limits Report -- 2.5 . Heat Flux Hot Channel Factot - Fo (X, Y, Z) (Specification 3/4.2.2) 2.5.1 Fj" - 2 . 32 2.5.2 K(Z) is provided in Figure 4 for Mark-BW fuel. e II 2.5.3 K(Z) is provided in Figure 5 for OFA fuel. The following parameters are required for core monitoring per the Surveillance Requirements of Specification 3/4.2.2: 2.5.4 (F8 (X, Y, Z) ]" = Fj (X, Y, Z)
- Mo ( X, Y , Z ) / ( UMT
- MT
- T ILT) where (FS(X, Y, Z) }" = cycle dependent maximum allowable design peaking factor which ensures that the Fo(X,Y,Z) limit will be preserved for operation within the LCO limits. (F$ (X, Y, Z) ]" includes allowances for calculational and measurement uncertainties.
F8 (X, Y, Z) = the design power distribution for Fg. F8(X,Y,Z) is provided in Table 1. F((X,1,Z) = the margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution. M o(X, Y, Z) is provided in Table 2. Note: (FS(X, Y, Z) ]" is the parameter identified as F7"(X,Y,Z) in DPC-NE-2011PA. 2.5.5 [FS(X, Y, Z) }"8 = F8(X,Y,Z) * (Mc ( X, Y, Z) / (UMT
- MT
- TILT) )
where [F$ (X, Y, Z) ]"5 = cycle dependent maximum allowab'le design peaking factor which ensures that the centerline fuel melt limit will be preserved for operation within the LCO limits. [ F$ (X, Y, Z) ]"' includes allowances for calculational and measurement uncertainties.
- F8 (X, Y, Z) = the design power distribution for F o. F8(X,Y,Z) is provided in Table 1.
g Mc (X, Y, Z) = the margin rema!.ning in core location X,Y,Z to the CFM limit in the transient power distribution. Mc (X, Y, Z) calculations parallel the Mo(X, Y, Z) calculation described in DPC-NE-20llPA, except that the LOCA limit is replaced with the CFM limit. F'< ( X , Y , Z ) is provided in Table 3. l __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________1_-_- -_
PRELIMINARY McGuire 1 Cycle 8 Core Operating Limits Report UMT - Measurement Uncertainty (UMT - 1.05). MT - Engineering hot channel factor (MT - 1.03). TILT = Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02, 2.5.6 KSLOPE - 0.045' where KSLOPE - Adjustment to the K value from OTAT 3 required to compensate for each 1% that Fa (X, Y, Z) exceeds its limit.
- typical value; actual values will be supplied when monitoring inputs are computed.
- - - - - - - - - - ~ _ _ - _ _ _ ___ __
FAWM!NM lMcGuire 1 Cycle 8 Core Operating -Limits Report - 1.2 (0,0,1.00) (8.0,1. 0 0) 1.0 (to8,o,94) 0.8 - N _ o,g - (12.0.0.647) M 0.4 - 0.2 -- I ;. 0.0 ' 0' 1 2- 3- 4- 5- 6 7 8 9 10 11 12 l Core- Height (FT.) o
- FIGURE 4 K(Z) - Normalized FO(X,Y,Z)_ as a Function of Core Height for Merk BW Fuel
PRELIMINARY
- McGuire .1 Cycle- 8 Core Operating Limits Report 1
1.4 - 4 1.2 - 1.0 0.8 - n N
\
0.6 - CORE HT.. (FEET) K(Z) 0.0 1,0000 6.0 1.0000 7.0 0.9396 0.4 - 10.0 0.500 9 10.8 0.9400 12.0 0.6470
- 0.2 -
0.0 0 -- 1 2 3 -4 5 6 7 8 9 10 11 .12 Core Height (FT.)
-FIGURE 5 K(Z) Normalized FO(X,Y,Z) as a . Function of Core Height for Optimized Fuel (Note: includes L(2) Penalty)
.=. .t
PREllMINARY McGuire 1 Cycle 8 Core Operating Limits Report 2.6 Nuclear Enthalov Rise Hot Channel Fa c t o r - Fa,, ( X , Y) ( Fin (X, Y ) ] * = MARP (X,Y) 2.6.1 McGuire 1 Cycle 8 Operating Limit Maximum Allowable Radial Peaks (MARP (X, Y ) ) are provided in Table 4. The following parameters are required for core monitoring per the surveillance requirements of 3/4.2.3: [ Fin ( X , Y ) ] "" = Fi, ( X , Y )
- MAH(X,Y)
Note: [Fia (X, Y) ]"" is the parameter identified as F""(X, Y) in DPC-NE-2011PA. where 2.6.2 Fin (X, Y) is provided in Table 5. 2.6.3 Ma, ( X, Y) is provided in Table 6. 2.6.4 RRH = 3.34 when 0.0 < P $ 1.0 where RRH = Thermal Power reduction required to compensate for each 1% that F 4 ,(X,Y) exceeds its limit. P = D ermal Power Rated Thermal Power 2.6.5 TRH = 0.01 where TRH = Reduction in OTAT K: setpoint required to compensate for each 1% that F,,(X,Y) 3 exceeds its limit. NOTE: Tables 5 and 6 will be supplied when monitoring inputs are computed. l
JREUMINAR1 Table 4. McGuire 1 Cycle 8 Operating Limit Maximum Allowablo Radial Peak 1.1 Axial Peg L2 Axial Peak 1.3 AxiaLEch E!evation (ft.L_ MARP MARP MARP 12.00 1.514 1.469 1 A28 11.67 1.523 1.482 1.440 11.00 1.540 1.507 1.464 10.33 1.553 1.533 1.491 9.67 1.564 1.558 1.518 9.00 1.572 1.578 1.549 8.33 1.579 1.597 1.500 7.67 1.583 1.607 1.607 7.00 1.588 1.615 1.631 6.33 1.591 1.621 1.645 5.67 1.599 1.627 1.656 5.00 t.597 1.633 1.665 4.33 1.599 1.637 1.672 3.67 1.601 1.641 1.679 3.00 1.602 1.644 1.683 2.33 1.604 1.647 1.687 1.67 1.605 1.649 1.691 1.00 1.606 1 651 1.695 0.33 1.606 1.652 1.698 0.01 1.606 1.652 1.698 1 A Axial Peak 1.5 Axial Peak 1.6 Axial Peak E!evation tft.1 MARP MARP MARP 12.00 1.388 1.350 1.314 11.67 1.399 1.361 1.325 11.00 1.422 1.382 1.346 10.33 1.450 1.408 1.371 9.67 1.479 1.437 1.399 9.00 1.507 1.464 1.731 8.33 1.536 1,491 1.448 7,67 1.565 1.519 1.473 7.00 1.593 1.547 1.500 6.33 1.621 1.577 1.531 5.67 1.649 1.606 1.559 5.00 1.677 1.633 1.583 4.33 1.696 1.660 1.608 3.67 1.710 1.687 1.634 3.00 1.718 1.712 1.659 2.33 1.724 1.736 1.685 1.67 1.731 1.756 1.711 1.00 1.736 1.771 1.735 0.33 1.737 1.774 1.754 0,01 1.738 1.775 1.760
l ELEl.lMINARX l Table 4. McGuire 1 Cycle 8 Operating Limit Maximum Allowable Radial Peak j (cont.) 1.7 Avial Peak 1.8 Avia! Pegg 19 Avial Peak E'evation f f t ) MARP MARP MARP 12.00 1.308 1.275 1.244 11.67 1.318 1.284 1.252 11.03 1.338 1.303 1.268 10.33 1.361 1.324 1.288 9.67 1.386 1.347 1.310 9.00 1.409 1.368 1.331 8.33 1.432 1.390 1.351 7.67 1.458 1.416 1.375 7.00 1.485 1.442 1.399 6.33 1.514 1.468 1.422 5.67 1.540 1.492 1.444 5.00 1.565 1.514 1.465 4.33 1.589 1.538 1.408 3.67 1.613 1.562 1.512 3.00 1.640 1.585 1.533 2.33 1.666 1.608 1.554 1.67 1.689 1.629 1.573 1.00 1.710 1.649 1.590 0.33 1.726 1.664 1.604 0.01 1.731 1.669 1.608 2.1 Aviat Peak - Eevation Ft 1 MARP 12.00 1.189 11.67 1.195 11.00 1.206 10.33 1.225 9.67 1.248 9.00 1.266 8.33 1.285 7.67 1.310 7.00 1.334 6.33 1 355 5.67 1.373 5.00 1.404 4.33 1.427 3.67 1.450 3.00 1.475 2.33 1.500 1.67 1.523 1.00 1.543 0 33 1.559 0.01 1.564 _ _____-_ _____-_- -__ - __ L _ -
L Attachment III FSAR Markups (To be incorporated in the 1992 FSAR update)
l l l for the purposes of this report, the following faults are included in this category:
- 1. Feeowater system malfunctions that result in a decrease in feedwater j temperature (Section 15.1.1).
- 2. Feedwater system malfunctions that result in an increase in feeowater flow (Section 15.1.2).
- 3. Excessive increase in secondary steam flow (Section 15.1.3).
- 4. Inadvertent opening of a steam generator relief or safety valve (Section 15.1.4).
- 5. Loss of external electrical load (Section 15.2.0).
- 6. Tureine trip (Section 15.2.3).
- 7. Inadvertent closure of main steam isolation valves (Section 15.2.4).
- 8. Loss of condenser vacuum and other events resulting in turoine trip (Section 15.2.5).
- 9. Loss of nonemergency AC power to the station auxiliaries (Section 15.2.6)
- 10. Loss of normal feeawater flow (Section 15.2.7).
- 11. Partial loss of forced reactor coolant flow (Section 15.3.1).
- 12. Uncontrolled rod cluster control assemoly bank withdrawal from a subcriti-cal or icw power startup condition (Section 15.4.1).
- 13. Uncontrolled rod cluster control assembly bank withdrawal at power (Sec-tion 15. 4. 2).
- 14. Rod cluster control assemoly misalignment (dropped full length assemoly, drooped full length assemoly bank, or statically misaligned full length assemoly) (Section 15.4.3).
- 15. Startuo of an inactive reactor coolant pump at an incorrect temperature (Section 15.4. 4).
- 16. Chemical ano Volume Control System malfunction that results in a cecrease in tne coren concentration in the reactor coolant (Section 15.4.6).
- 17. Inaovertent operation of the Emergency Core Cooling System ouring power
. operation (Section 15.5.1).
- 18. Chemical ano Volume Control System malfunction that increases reactor coolant inventory (Section 15.5.2).
- 19. Inadvertent opening of a pressurizer safety or relief valve (Section 15.6.1),
- 20. $ rea k u n s n sieu men t inn e or okher lunes From reackor coohnt fressure sou .,da ry +h a+ (e n e 4, ra + e. Co niornment (Sec4 eon (5. 4 2. ).
15.0-3 1985 Update
! 3. ' Reactor coolant pumo shaf t seizure (locked -rotor) (Section 15.3.3).
. 14 Reactor coolant pumo shaf t break (Section 15.3.4). < 5. Spectrum of red cluster control assembly ejection accidents (Section 15.4.8). ,
s' 6.. Steam generator tube failure (Section 15. p )
- 7. Loss of coolant accidents resulting from the spectrum of postulated piping g- breaks within the reactor coolant pressure boundary (large break) (Section
- 15. 6.g).)
\5) '
- 8. Desigifbasis fuel handling accidents (Section 15.7.4).
.15.0.2 OPTIMIZATION OF CONTROL SYSTEMS A control system setpoint study is performed in order to simulate perfor-mance of the reactor control and protection systems. In this study, emphasis is placed on the development of a control system which will automatically
, . maintain prescribed conditions in the plant even under a conservative set of l reactivity parameters with respect to both system stability and transient l performance. p For each mode of plant operation, a group of optimum controller setpoints l is determined. In areas wnere-the resultant setpoints are different, com-I promises _ cased on the optimum overall performance are made and verified. A consistent set of control system parameters-is derived satisfying plant opera- - l tional requirements throughout the core life and for various levels of power l operation. l =The study comprises an analysis of the following control systems: rod control, l- steam dump, steam _ generator level, pressurizer pressure and pressurizer level. 15.0.3 PLANT CHARACTERISTICS AND INITIAL CONDITIONS ASSUMED IN THE ACCIDENT. ANALYSES 15.0.3.1 Power Ratinos Table 15.0.3- lists the Prkeipa4 power rating values which are assumed in analyses performed in this. report. Two ratings are 9 4wen: con ssdered
- 1. The guaranteed HSSS thermal power _outout, This power output includes the thermal pcwer generated by the reactor coolant pumps, l- 2. The engineered safety features design rating. The NSSS supplied engineered safety features are designed for thermal. power higher than the guaranteed value in order not-to preclude realization of future potential power-capability. This higher thermal power value is designated as the engineered safety features design rating.' This power output includes __
the thermal power generated by the reactor coolant pumps. Allowance for errors in the determination of the steady-state power level is made as described in Section 15.0.3.2. 15.0-5 1984 Update
Mhem:Lp0wer v:lu : u::d fer :::n-tran:icnt 2nalyzed ere giver " T:ble --15.0.34 In all cases wnere the engineered safety features design rating is used in an analysis, the resulting transients and consequences are conserva-tive comparea to using the guaranteea NSS5 thermal power output rating, w = cf other acetin. a A s pere = r: uti m ed in tne m ie:nt sn e - 595 170 giV n i7 Ie' wle 10. 0. 3 C. 15.0.3.2 Initial Conditions . For most accidents which are DNB limited, nominal values of initial conditions are assumed. The allowances on power, temperature, and pressure are determined 4 basis and arg on a statigi * "Wocedures:5[ncluded Referengesl kar.sr. in the "I";reved c: the limit DNBR, The -as described
!! Oc;ign in IrcccUw c,"
and is sed more fully in Section 4.4.
,.Qr sta4ishcal Core Dess&
For accident p hlkn are not DNB limitec, or for which the Improved Thermal DesignvFrocedures J nct employed, the initial conditions are obtained by adding the maximum steady state errors to rated values. The following conservative steady state errors were assumed in the analysis:
- 1.1 % allo wa nc e. &r msesurement errar}
1t4bcioe' coolan+ loof >
- 1. Lore power 12 percent allowance for calorimetric error
- 2. Average Reactor Coolant 4*F allowance for controller System temperature deadba'nd and measurement error
- 3. Pressurizer pressure +30 -42 pcunds per square inch (psi) g allowance for steady state fluctua- y tions and measurement error g The control band allowances for pressurizer pressure include the effect of the b observed thermal non-repeatability of ITT-Barton class 1E transmitters used at McGuire. <$
s+a lis4ica t Table 15.0.3-3 summarizes the mputer codes used in the accident analyses and shows which accidents employed a NB analysis.u @ he Improved h em:1 0 :ign tsceuwi c. 15.0.3.3 Power Distribution The transient response of the reactor system is deoendent on the initial power cistribution. The nuclear design of the reactor core minimizes aaverse power distributions through operating instructions and the placement of control rods. Power distribution may be characterized by the radial factor (Fg ) and the
;otal ceaKing factor (F q). The peaking factor limits are given in tne Tecnni-cal Specifications.
For transients which may be DNB limitec, the radial peaking f actor is of impor-tance. The racial peaking f actor increases with decreasing power level due to rod insertion. This increase in F y is included in the core limits illustra-tea in Figure 15.0.3-1. All trancients that may be DNB limited are assumed to 15.0-6 1956 heu
i l l begin eith a F g consistent with the initial power level defined in the Tecn-nical Specifications. The axial power shapes used in the DNB calculation are discusseo in Section 4.4 The radial and axial power aistributions described above are input to the 4HiNC- code as cescribec in Section 4.4. VITR E .o g For transients which may be overpower limited, the total peaking factor (F ) 9 is of importance. All transients that may be overpower limited are assumed to begin with plant conditions, including power distributions, which are consistent with reactor operation as defined in the Technical Specifications. Some overpower transients are slow with respect to the fuel rod thermal time constant, for example, the Chemical and Volume Control System malfunction that results in a decrease in the boron concentration in the reactor coolant incident, which lasts many minutes, and the excessive increase in secondary steam flow incident, which may reach equilibrium without causing c reactor trip. For these transients the fuel rod thermal evaluations are performed as discussed in Section 4.4, Other overpower transients are fast with respect to the fuel rud thermal time constant, for example, the uncontrolled rod cluster control assembly bank withdrawal from suberitical or low power startup and the rod cluster control assembly ejection incidents, which result in a large power rise over a few seconas. For these transients a detailed fuel heat transfer calculation must be performed. Although the fuel rod thermal time constant is a function of system conditions, fuel burnup, and rod power, a typical value at beginning-of-life for high power rods is approximately 5 seconds. 15.0.4 REACTIVITY COEFFICIENTS ASSUMED IN THE ACCIDENT ANALYSES The transient response of the reactor system is dependent on reactivity feed-back effects, in particular the moderator tem;1erature coefficient and the Dop-pler power coefficient. These reactivity coefficients and their values are dis-cussed in detail in Chapter 4. In the analysis of certain events, conservatism requires the use of large reactivity coefficient values, whereas in the analysis of other events, con-servatism requires the use of small reactivity coefficient values. Some events, sucn as the loss of reactor coolant from steam generator tube ruptures, do not depend on reactivity feedback effects. The values assumed for these coefficients are given in Table 15.0.3-4 Reference is made in that table to Figure 15.0.4-1, which shows, as a function of power, the upper and lower bound Doppler power coefficients used in the transient analysis. The justification for the use of conservatively large versus small reactivity coefficient values is treated on an event-ey-event basis. In some cases conservative comoinations of carameters are used to bound the effects of core life, althougn these combinations may not represent possible realisti: situations. 15.0.5 ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS The negative reactivity insertion following a reactor trio is a function o+ "e position versus time cnaracteristic of the rod cluster control assemblies ad of the variation in roc wnrth as a function of rod position. With respect to accident analyses, the critical parameter is the time of insertion up to the dasnoot entry, approximately 85 percent of the rod cluster travel. The rod l l 15.0-7 1984 Codate
. . . J
, , cdsrenad a *d b cluster control' as'semoly position versus time (, characteristic assumed in the accident analyses is shown in Figure +15.0.5-17 The roc cluster control assemoly insertion timeMo dashoot entry. 4+-teen :: 2.3 :ecenca. Ret 1 Brop times are only shply deoenaent on the type of rod cluster control assemolies actually used in the plant, whether Ag-In-Cd (Unit 1) or B,C (Unit 2). Mcwever, ac444ents-ar4--
+omeevaMvdy--analym usW the !cnger-deep time. Table 15.0.5-1 gives the (,4 drop times assumed for each FSAR analysis. m eoW Figure 315.0.5-Mshow)( the fraction of total negative reactivity insertion versus normalized rod position for a core where the axial distribution is skewed to the lower region of the core. An axial distribution which is skewed to the lower region of the core can arise from an unbalanced xenon distribution. This curve is used to compute the negative reactivity insertion versus time following a reactor trip, which is input to all point kinetics core models used in transient analyses. The bottom skewed power distribution itself is not input into the point kinetics core model. There is inherent conservatism in the use of Figures 3.0.5-2 in that is based on a skewed flux distribution which would exist reTTf.ively infrequently. For cases other than those associated with unbalanced xenon distributions, signi-ficant negative reactivity would have been inserted due to the more f avorable axial distribution existing prior to trip. a and b The normalized rod cluster controVassembly Aegative reactivity insertion versus rve5thown in this figure 7dd'obtained time is shown from Figures in figures 15.0.5-2. 15.0.5-1_and 15.0.5-3F AThepp"4M" Fe negative reactivity insertion,fol-lowing a trip (of 4 percer.t ak/kJis assumed in the transient analyses except where specifically noted otherwise. This assumption is conservative with respect to the calculated trip reactivity worth available as shown in Table 4.3.2-3. For . Figures 15.0.5-1 and 15.0.5-2, the rod cluster control assembly drop time is nor-malized to 3.3 seconds. ihe m/ve- used in the en=Irses.
<n o.h The normalizea frod cluster control assemoly negative reactivityVnsertion versus /
time curve'for an axial power distribution skewed to the bottomt (Figures 15.0.5-3( is used in those transient anal o is usec. I.svt:. -Crs, - cAthc.ke.d pogo. yses for which a point kinetics c~ re mocel Where special analyses require use of three dimensional +r-ancl On dimen&m! core models, the negative reactivity insertion resulting from the reactor trip is calculated directly by the reactor kinetics code and is not separable frcm the other reactivity feedback effects. In this case, the rod cluster control assemoly position versus time of Figure 15.0.5-1 is used as code input. 15.0.6 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES A reactor trip signal acts to open two trip breakers connected in series feed-ing power to the contro-l rod drive mechanisms. The loss of power to the mech-anism coils causes the mechanisms to release the rec cluster control assemo-lies which then fall by gravity into the core. There are various instrumenta-tion celays associated with each trip function, including delays in signal , actuation, in opening the trip breakers, and in the release of the rods by the mecnanisms. The total delay to trip is defined as the time delay from the time P;at trip concitions are reached to the time the rocs are free and begin to fall. Limiting trip setpoints assumed in accident analyses and the time delay . assumed for eacn trip function are given in Table 15.0.5-1. Reference is maae in that taole to tne overtemperature and overpower aT trips snown in Figure 15.0.3-1. I 15.0-8 1986 Update
Transienta analyzed-with the LOFTRAN and-NOTRUMP codes use the RCCA insortion figures denoted by "a", while the RETRAN-02 analyses employ the."b" curves. l-l .' l l
This figure cresents the allowable Reactor Coolant 1000 average temperature ano aT for the aesign flow ano cower distrioution as a function of crimary coolant pressure. The councaries of coeration cefinea ey tne overcover aT trip and the overtemoerature aT trip are representea as " protection lines on this diagram. The protection lines are arawn to include all adverse instrt.tentation and setpoint errors so that uncer nominal conoitions trip would occur well within the area bounced by these lines. The utility of this diagram is in tne fact that the limit imposea by any given DNBR can be represented as a line. The ONB lines reoresent the locus of conditions for wnich the ONBR equals the limit value 41-P 'r N t*c!: ::l' r: 17.m ...c .y ; o 1 .c i l . 7 All points below and to the lef t of a ONB line for a given pressure have a ONBR greater than the limit value. The diagram snows that DNB is prevented for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable DNBR line at any point. The area of permissible operation (power, pressure and temoerature) is bounced by the comoination of reactor trios: high neutron flux (fixea setooint), high oressure (fixea setpoint), low pressure (fixea setooint), overoower and overtemoerature 2T (variaole setooints).
>+aisRed desia n p roc edure s Thelimit[value,wnichwas'usedastheONBRlimitforallaccidentsanalyzed wi th -tee- hg eveu Th m oi eng, M c cure (see Table 15.0.3-3), is conservative compared to the actual design DNBR value G.:1 im ...e ......~.c sci w e 1. 0 3 --
'r M: ypicC c!' required to meet the DNB design oasis.
The difference between the limiting trio point assumed for tne analysis and the , nominal trip point represents an allowance for instrumentation cnannel error and setpoint error. Nominal trip setooints are specified in the clant Technical Specifications. During plant startuo tests, it was cemonstrated that actual instrument time celays are eaual to or less than the assumea values. Addition-ally, protection system cnannels are calibrated ana instrument response times pg are ceterminea periccically in accoraance with the Tecnnical 5cecifications, f 1.c 7, INSTRUMENTATION ORIFT AND CALORIMETRIC ERRORS 20WER RANGE NEUTRON FLUX The instrumentation 4 't and calorimetric errors usea in estr' sning the oower range hign neutron , setcoint are presentea in " e 15.0.7-1. f;. I The calorimetric error is the error a. 'med i- . e cetermination of core ther-mal cower as obtainea f rom seconaary ol- - asurements. %e total i n cnam-cer current (sum of tne t:0 anc c~ . section is :3iioratec (set eausi) to this measurec power on a per' c oasis.
'e secondary cow
. .[cotainea'ommeasurementsof'eecwate ' low, eeawater inlet tempe e to tne steam generators, ano steam cressure. % accuracy instrur ation is crovidea for tnese measurements with accuracy toler es
~ - ighter than tnose wnicn woula be requireo to control 'esawater flow.
'5.0-9 .9% cpaate
W 15.0.8 OPERABILITY OF NON-5AFE1Y GRADE EQUIPMENT (
)
The transient analyses presented in Chapter 15 only assume non-safety grade ( systems and equipment are operable in the following situatio'ns: ) (
- 1) When operation of the system will cause the transient to be more severe. )
If there is doubt about the system's effect, the transient is analyzed 1 ( and presented with and without the system available. ) (
- 2) When a loss of a non-safety grade system initiates a transient by itself, )
it is not superimposed upon other transients unless there is a credible ( reason that one would cause the other. The following non-safety grade systems and equipment are assumed operable in ( some analyses presented in Chapter 15.
- 1) Automatic rod control. )
huAus (
- 2) Pressurizer pressure control (power operated relief valvesyand spray). )
Main Feedwater System. (
- 3) )
AUTOMATIC ROD CONTROL ( The Automatic Rod Control System is assumed to be operable in the following ( transientanalys{s: ) 15.1.3 Excessive increase in secondary steam flow. (
)
15.4.3 Rod Cluster Control Assembly Misoperation. ( 15.6.1 Inadvertent opening of a pressurizer safety or relief valve. ( Analy ps of the excessive increase in secondary steam flow transient are done with ano without automatic rod control and are presented in Section 15.1.3. ( Both cases meet acceptance criteria. For the dropped rod (s) case of Section 15.4.3 the automatic rod control system is assumed to be operable. Because of the power overshoot possibility associated with automatic control, it is conservative to assume its function for this accidert. For the inadvertent opening of a pressurizer safety or relief valve transient it is conservative to assume the automatic rod control system is operable. ( , During this transient RCS pressure will be decreasing. Decreasing RCS , pressure will cause the reactor power to decrease due to moderator density ( feddback. If the Automatic Rod Control System is operable it will function to / maintain power and average coolant temperature, thus causing a more severe N transient, d l l i 15.0-10 12/37 __x_____ _____ _____ __ _ - _ A
- 4E55URIZER ORE 55URE CONTROL IDower oceratea Relief Valves ana 5crav) !
I
'Se ; essuri:er Pressure Controi ifstem is assumea to be operarie '- tne I
'ollowing transients: ._
15.c. Turoine trio. Q t f/C c t. 15.4.2 Uncontrois s/,/ i I i j quster control assemoly banx-witfarawal st power. t r1 S C P & Analysis of the turoine trio event wi .,
/
41thout pressurizer cressure fro m control is presented in Section 15.2dof the
. Both cases naet scceptance criteria.
ggc/1c s( . f9 O For the rod bank MthdIwal transient it is conservative to assum pressure ca 6fis operaole. The limiting criteria for this transien essurizer the
; ONB r . Maintaining RCS pressure low will result in lower DNB rstics.
MAIN FEEDWATER CONTROL S~fSTEM The Main Feeawater Control System is assumed to function curing the following transient analyses: 15.1.1 Feeawater System malfunctions causing a reauction in feeawater temperature. 15.1.2 Feedwater System malfunction causing an increase in feeawater flow. 15.1.3 Excessive increase in seconaary s ?am flow. 15.1.5 Steam system oipi.,g f ailure.
'5.4.2
. 'Jncontrollea roc c!uster control assemoly cank withdrawai at power.
'. 5 . 4 . 4 5tartuo of an inactive loco.
'.5. 5.1 Inaavertent. oceration of Emergency Core Cooling System curmg ::ower
- ceration.
ii
'5.6.1
. :naavertent ocening of a cressurizer safety or relief salve.
;oss of main feeawater flow is a Conaition II occurrence oy itself anc is !
analyzea ', 5ecti on 15. 2. 7. 5ere is no creaiole "eason 'or any 9 : e
- .:natt:en .: events i stea a::ove to cause a icss of feeawater ':w.
Therefore. a loss of f eeawater is not consicereo coincicently witn nose ! occurances listea above wnicn are Concition II. !
- or tne steamline creak transient it i s conservative to assume main 'eeawater ;
is avaiisole. This maximizes tne amount of steam generator inventerv i availaole to ce n': u , anc crolongs tne transient, j blown down '
.: q - ;; g 2-oe 2.,- .,4e-, -- - : , _ g _.4 - - n --n n- -iw. ,,1 e: 2; - l curo: :n .t : -;:1: 2 .:.1 :. - I
.=..n.,,
.-.--.o
15.2.3 Turbine trip 15.3.1 Partial loss of forced reactor coolant ficw 15.3.2 Complete loss of forced reactor coolant flow 15.3.3 Reactor coolant pump shaft seizure (locked rotor) 15.4.1 Uncontrolled RCCA bank withdrawal from a subcritical ,r low power startup condition 15.4.2 Uncontrolled RCCA bank withdrawal at power 15.4.3d Single RCCA withdrawal at full power DNB limited transients are analyzed assuming that pressurizer pressure control is available to maintain pressure low since this is conservative. DNB transients which might cause RCS pressurization are also evaluated with pressurizer pressure control unavailable. Since a late reactor trip generally reduces the margin to the acceptance criteria in any transient, the effect of operable pressurizer pressure contre. in delaying reactor trip is considered for those transients where it would occur. _a. . _ . _ _ _ . _ _ _ - .___ -
1 in a parhr core Ergt-CELL ba sed PDQ-XD model. PDo-XD l c.alevlafeo a v2 rage Zis sion raies over cach burnup c+ep for ail 4he sq ruheed tsa+qes o+ Pa and u. /
\
I 15.0.9 FISSICN PRODUCT INVENTORIES i 15.0.9.1 Insenterv in the Core Fission rates and total fissions are determined ec?the 'ollc*g W 4cr m en5. r uci desir i h* nn ie
~
m ed ta u ccl t": cactr are ae i u v u a ;. i j nenc.sn c.2 .l0 Pair ia a rE'lT (%ftW7 ice 4 ) , 4 <ciaivo vi alcukt: ' n tanten m . 2nc svac;ge ;e-cc f#eme ;aww.vua D. Atwt is ouu w a2 4 iunLLiva vi wi nus .nd pencr 2cus i sj . . . a4i Ur st.M-Mj sinni" ant is nowu= vi w i u d ai um ena ui oni ci. This data along with appro-priate power levels, burnuos, peaking factors and isotope pnysics information is then input into a Duke coce, FISSION, anicn ca.culates the fission rates and total fissions occurrirg (by isotope), for either a single assemoly or a three region core, at the instant before shutdown. Total core inventories for all isotopes except the long livec Kr-85 are determined by summing the products of the fission ratet and the fission yields for eacn of the fuels consicerec (U-235, U-238, Pu-241). The inventories tnus cotainea are the eouilibrium valves. Because of the long nalf life of Kr-85 Therefore, its (10.76 years), it coes not acnieve an equilibrium builduo. inventory is obtained from the total fissions which have occurred at the ena of core life and tne fission yields for eacn of the fuels. 15.0.9.2 inventerv in the Fuel Pellet Clad Gao The raniation sources associated with accidents whien may cause more than 1 percent failed fuel (loss of coolant accident, rod cluster control assemoly ejection, ano fuel handing accidents) are cased on the assumotion that the fission products in the gap between the fuel pellets and the claccing of tne d6magea fuel roos are released as a result of cladding f ailure. The gap activitiu were aetermined using the mocel suggestcc in Regulatory Guice 1.25. Specifically,10 percent of tne iocine ana noole gas activity (except sc-85 rnicn is 30 cercent) is accumulated in the fuel-claa gao. 'he
- ore ano associateo gap inventories usea for the LOCA anc roa ejection sccicents are snown in Table 15.0.9-1.
15.0.9.3 Inventories in the Reactor Coolant The raciation sources associated with accicents wnicn are not expected to cause fuel failures (main steam line creak. loss of external 'oaa, steam
;enerator tuce ructure, ana liauid waste tanx ructure) are taKen f-om Taole 7-7 of Reference 4 for iccine anc nooie gas '3noces. 'ese wentories are ,
given in Table 15.0.9-2.
;5.0.10 RESIOUAL CECAY WEAT
/ .T n s e r& Srem a b ci, eri pay e.
Total Residual Weat s, 15.0.10.1 s-..- x Residual heat in a subcritical core is calculated for tne loss of coolant acticent oer tne requirements of Accencix s of 10 CFR 50 > C.as cescribea in
- References 5 ana 6. These recuirements incluce assuming infinite irradiation
* .ime oefore the core goes succritical to cetermine tission crocuct dec3y '
T-er V ' m c; u wents. ,. m u.o c ;ce ; am 5ec oss u . .n 2 2 .vu eneroy. o rnau ~ nca; rp . ;ane-e m er ne esau n.; .: :: w u. e il4 7,.- m1 9 (- M 15.0-12
i For transients analyzed with the LOFTRAN code, the same models are used except that fission product decay energy is based on the core average exposure at the end of an equilibrium cycle. For transients analyzed with the RETRAN-02 code, f.ssion product decay energy is also based on the core average exposure at the end of an equilibrium cycle. The model used in the RETRAN-02 analyses is based on Reference 10, with a two standard deviatien uncertainty adjustment applied to the results in lieu of the 1.2 factor of 10 CFR 50 Appendix K.
. _ _ _ _ _ _ _ _ _ _ -_w
'.i . 0.10. 2 Distribution of Cecav " eat :nilowina toss of C:olant acc1 cent fA During a loss of coolant acc10ent. .ne core i s rapidly snut cown oy void f or-mation, rea cluster control assemoiy insertion, or both, and a large fraction of the neat generation to be c:nsicereo comes from fission proauct cecay gamma rays. This neat is not cistricutea in the same manner as steacy state fission power. '.ocal peaking effects onicn are imoortant for the neutron depenaent part of the neat generation do not aoply to the gamma ray contricution. The steacy state f actor of 97.4 percent wnien reoresents the fraction of heat generated within the clad and cellet drops to 95 percent for the not roa in a loss of coolant accident.
For example, consider the transient resulting from the postulated double enaea break of the largest Reactor C0olant System oipe. 1/2 secono after the rupture aoout 30 cercent of the heat generated in tne fuel rods is from gamma ray absoro-tion. The gamma oower shape is less peakea than the steaay state fission power snace, reaucing the energy cecositea in the not roa at the expense of adjacent
- cider reas. A conservative estimati of this effect is a reauction of 10 cercent of the gamma ray contribution or 3 cercent of the total. Therefore the available neat cepositea in the not roc at the expense of the other roos is reauced to 97 percent of its former value. Since the water density is consiceraoly reaucea at this time. an average of 98 percent of this available heat is cecosited in the clad and cellet, the remaining 2 percent being absoroed oy water, thimoles, sleeves anc grics. The net effect is a factor of 0.95 rather than 0.974, to be appliea to the neat proauction in the not rod.
15.0.11 COMPUTER CODES UTILIZED $ kmaaries of some of the princicai ::mouter coces usea in transient anaiyses are given celow. Other coces, in ; articular very soecializea coces in wnicn the moceling nas oeen cevelocea to simulate one carticular acciaent, sucn as those asea in tne analysis of the Reactor C0olant System oipe ructure (iection 15.5.4), are summar1:ec in their respective accicent analyses sections. 5e ccaes usea in the anaiyses of eacn transient nave oeen listed in Taole 15.0.3-3. RM _ 15_0. 11.1 FACTRAN
* . /,/ k od ::CTRAN calculates the transient tamcerature cistricution in a cross seckon
/
D g of a metal clacWfuel roa ano tne transient neat flux at tne sartace of tne
- ad using as inout tile nuclear cower anc the time-cecencent c:Mant carameters {
':ressure. Mow, temoerature, Inc :insity). ~5e c:ce uses ,3-wei nocel enicr exnibits tne f 0llowing f eatures simuitaneously: '
- 1. A sufficiently large numoer of taial soace-in'crements to nanale 4st -
transients sucn as roa ejecucn ac:1 cess.
,/
- . Material crocerties wnicn acV0nctions of temoerature ano a 5:anisti-catea f uei-to-claa gap ed t-ansfer calculation.
? The necessaryA culations to nanale cost-ONB transients: 'ilm coil-ing r. eat,trEnsfer correlatlans. :ircaioy water reaction ana cartlai ,
me ttiM of tne materia ls. I
/
pCfRAN 13 furtner aiscusseo i~ Aeference 7. p
'5.0-13 U
______________________1__ _ _____ _
. _ - - . . - _ _ . . . . _ . _ . - . . - ~ _.__--.--.- -~ _ . .
l
-1 1
-15.0.11.1 RETRAN-02 RETRAN-02 is a code capable of simulating most thermal-hydraulic l transients of interest in both PWRs and-BWRs. It has the flexibility to model any general fluid system by partitioning the system into a one-dimensional network of fluid volumes and connecting flowpaths or l junctions. The mass, momentum, and energy conservation equations are l then solved by employing a semi-implicit solution technique. The time step selection logic is based on algorithms that detect rapid changes- 1 in-physical processes and limit time steps to ensure accuracy and I
~ stability. Although the equations describe homogeneous equilibrium fluid volumes, phase separation can be modeled by separated bubble rise volumes and'by a dynamic slip model. The pressurizer and other '
regions can be modeled as non-equilibrium volumes when such phenomena are present. Reactor power generation can be represented by either explicit input as a function of time, a point kinetics model, or a one-dimensional kinetics model. Heat transfer across steam generator tubes and to or from structural components can be modeled. Special component models for centrifugal pumps, valves, trip logic, control ; systems, and other features useful for fluid system modeling are available. RETRAN-02 is further discussed in Reference 11. l... L I l l
j 15.0.11.2 LOFTRAN The LOFTRAN orogram is usea f:r stuaies or transient response of a FWR sjstem to specifiea certuroations in crocess parameters. LOFTRAN simulates a multi-loop system oy a mocel containing reactor vessel, not anc cold leg pioing, steam generator (tuce ano snell sices) anc the cressurizer. The cressurizer neaters, spray, relief ano safety valves are also consiaereo in the program. Point mocel neutron xinetics, ana reactivity effects of the mocerator, fuel, Doron ana roas are incluaed. The seconaary siae of the steam generator utilizes a nomogeneous, saturatea mixture for the thermal transients ano a water level correlation for inoication and control. The Reactor Protection System is simulated to incluae reactor trios on hign neutron flux, overtemoerature .iT, overpower aT. nign ana low pressurizer pressure, low flow, and nign pressurizer level. Control systems are also simulated including roa control, steam cumo, feeowater control ano pressurizer pressure control. The Emergency Core Cooling. System, inclucing the accumulators anc upper neaa injection, is also modelea. LOFTRAN is a versatile program wnicn is suitea to botn acciaent evaluat1on ano controi stuoies as well as carameter sizing.
-LOFTRAN also nas the capability of calculating the transient value of DNBR baseo on the input from the core limits illustrated on Ficure 15.0.3-1. The core limits represent tne minimum value of DNBR as calculatea for typical or thimole cell.
LOFTRAN is further discussed in Reference 8.
! . 11.3 TVINKLE The TWINKL: ogram is a multi-cimensional soatial neutron kinetics c . wnien was patternea a .e.r steaay state coces cresently useo f r reactor,,etre cesign, g plsce The ceae uses an imahc.i{ finite-aifference metnoa to solve Ine'two grouc tran-4 get' g sient neutron ciffusion-ecua.ttons in one, two ano tnree almensions. the c:ce uses six ceiayeo neutron grouoba.qa contains a aetaH1ia multi-region fuei-claa-
~
1 47 4 cacK coolant neat transfer moael for. calh4 ting pointwise Ococier ana moaerator 'eea-effects. The coae nanales uo to 20E5patial-points, anc cerforms its own
- @c'h . steaay state initialization. - Asice fcom'casi ross-section cata anc tnereal-p*3 nyaraulic parameters, tne coce accepts as incut c ~ c : riving. functions suen as ,
inlet temperature, pressure.,Jhfw, coron concentratioNc:ntrol roa matten, ana others. Various edits arstroviaed, e.g. :nannelwise cow axial offset, entnal-cy, volumetric surgp-pointwise cower, anc fuel temoeratures.
\
- ne xinetic enavi: of a react:?-f:r tran-
~
The TWINKLE.Co~ce is usea to creaic i 3ients .nica cause a major certurcation in ne scattai eutron flux c:str',out':n.
'- N TWI.N_KLE- _- . is furtner cescribea in Reference 9.
\
M41. * ~5 aHer Secieon 15.0.(, Th: 2'": O= . = =ri=c " k nior [o" h)
- 15. 0. M ENVIRONMENTAL CONSEOUENCES.
7 D a summary cf ; e of f site :.ses s cresentec ', 's o l e _5. . . ,)it -.
- f eacn ac::icent analysis is g ven in tne accro:riate section.) A cescr*: . n
( wkh close. 6%s even poca4hd8cdly) ( ..
- 15. c- 9
15.0.1143' VIPRE-01 VIPRE-01 is a subchannel thermal-hydraulic computer code. .With a subchannel analysis approach, the nuclear fuel element is divided into
.a number of-quasi one-dimensional channels that communicate laterally
- by' diversion crossflow and turbulent mixing. However, VIPRE-01 is
- also capable of simulating single subchannel geometry. Given.the l
geometry of the reactor core and coolant channel, and the boundary conditions or forcing functions, it calculates core flow distributions, coolant thermodynamic conditions, fuel rod tempera-tures,- and the departure from nucleate boiling ratio (DNBR) for l steady-state and transient conditions. VIPRE-01 accepts all necessary boundary conditions originating from a system transient simulation code or transient core neutronics simulation code. Included is the capability to impose different boundary conditions on different i segments of the core model. For example, different transient inlet , - temperatures,. flow rates, heat flux transients, and even different l- transient assembly and rod radial powers or axial flux shapes, can be modeled. VIPRE-01 is further discussed in Reference 12. 15.0.11.4 . NODE-P NODE-P is a multi-dimensional nodal code which computes the core effective multiplication factor, the three-dimensional core power distribution, core coolant flow and temperature distribution, and fuel exposure distribution. The program includes the effects of partially inserted full-length control rods and up to 25 different fuel assembly types with different enrichments and burnable absorber shim loadings.
' The program iterates to account for the interaction between power distribution and core nuclear properties which-depend on coolant flow and coolant temperature distributions, fuel temperature distribution, and xenon distribution. It computes the time dependence of xenon following changes in power level and/or change in power distribution.
The_ program permits fuel shuffling from one location to another and -
-fresh fuel insertion for burnup cycle calculations. Individual steps can be stacked for either xenon transient or fuel cycle burnup-calculations. NODE-P is used.as the primary reactor simulator in the i
-generation of safety analysis physics-parameters, maneuvering L -analysis, and startup and operation reports.
L l NODE-P is further discussed in Reference 9. i t
- - -- - - .-,u 3 -, -
e
l 15.0.11.5 PDQ solves the neutron PDQ is a fine mesh diffusion theory code two, thator three dimensions in up to diffusion equation in either one,PDQ also performs a depletion calculation This using four energy groups. for cross-section variations. an interpolation scheme to representation flexible account of time-dependent cross-scheme provides a for specification of nuclide chains. oups. sections and a general format core and PDQ is typically run in two dimensions Results from PDQ and two energy grExplic four quarter assembly (color set) geometries. l tions. provide the_ basis for all neutronics codes and calcu a PDQ is further discussed in Reference 9.
1 i REFERENCES FOR SECTION 15.0
- 1. Chelemer, H. , Boman, L. H. , Sharp, D. R. , " Improved Thermal Design Procedures", WCAP-8567, July, 1975.
i
- 2. "0CELOT - A Spectrum Dependent Non-Spatial Depletion Code", S. M. Stoller i Corp. ,
1
- 3. Barry, R. F., " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094", WCAP-3269-26, September, 1963.
]
l 4 Radiation Analysis Design Manual - William B. McGuire Units 1 & 2, j Westingnouse Electric Corporation, Revision 3, January, 1976.
- 5. Bordelon, F. M. , et al, " SATAN-VI Program: Comprehensive Space - Time Dependent Analysis of loss of Coolant",-WCAP-8302 (Proprietary) and i WCAP-8206 (Non-Proprietary), June 1974. l
- 6. Bordelon, F. M., et al, "LOCTA-IV Program: Loss of Coolant Transient Analysis", WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.
~
,h/,"'..k:A d h 1 ".ht""
-__, . _ . _ , ...m
- 8. Burnett, T. W. T. , McIntyre, C. J. , Buker, J. C. , Rose, R. P. , "LOFTRAN 1 Code Description", WCAP-7907, June,1972.
-? *:,nc, ,s._,o, ei . , 5m i j , a. . . ' " T " 'l ' "" ' + ' - C ' - - > ' . m . vn
. ,_ L L.;._:::
^
_::" ~ : '" '"
, y , ~:?" 2 " -^ '-~ 'm.
bie.d dro- % d e ck g o 15.0-15 12/ S T
. . ._- . . . - . _ _ - _ . .. . - . . . ~ . - . . .._
- 7. L" Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01", DPC-NE-2004P, Rovision 1, February 1990.
- 9. Nuclear Physics Methodology for Reload Design, DPC-NF-2010A, Duke Power Company, May 1985.
10, American Nuclear Society, " Decay Heat Power in Light Water Reactors", 7NST/ANS 5.1, 1979.
- 11. EPRI, "RETRAN-02: A Program for Transient Thermal-Hydraulic Analysis of Complex' Fluid Flow Systems", EPRI NP-1850-CCM,.
Revision 4, November 1988.
- 12. EPRI, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores",
EPRI NP-2 511-CCM- A, Revision 2, July 1985. 4
- ~
.t.
f /
./
s Table 15.0.3-1 s Nuclear Steam Sucoly Svstem Power Ratinas / x\, / Transients Analvzed at 0 W t
/
/
0 Hot Zaro Tgrmal Power N Reactor Coolaht.Dueo Thermal Power (assumed) 0 N Transients Analvted at 3425 Wt Guaranteed Core Thermal Power / 3411 Reactor Coolant Pumo Thermal Rower N 14 N
\ '
Transients Analyzed at 3493 Wt Guaranteed Core Thermal Power (/2 percertt x 3479 Reactor Coolant Pumo Therma) ower \ 14 x
'N Transients Analyzed at 36B0 Wt
/
Engineering Safety /eatures Design Rating (+2 pereen 3630 (maximum calculated turoine rating)
/ \
- Reactor Coolant' Pumo Thermal Power 'N 20 j x
, Transients Analvzed at 3479 Wt l Guaran/ teed Core Thermal Power (+ 2 percent)
\
347 React'or Coolant Pumo Thermal Power O j s
\ i
- s i x
l f.
'.2 / 3 7 L
l l
~ ,
11II ji! e t s _ o - p U C 0 9 4 s d r a u g e rn r f ee S a ii d tt e rr l y m z y ee s s a w 32 r e u m l a nn a r nnrh ooep i i
- i x n k d iiti a A r y n h ggnr rrrr M t t ) adt eeee eeee s
e s e i i v v ( sboi rw RR cp wwww oooo AAA , CCC4686 d s i i dd *+nn pppp OOO . o a t t oodd *+io LLL0000 C r c c rreo l ol o BBB = = = = a s nr 1I22 l rl r SSS _ s f e c ddg ueue CCCC e o r r eeie nnnn f zf z hhh t u y t . ppl ppag l oooo cccGGGG
. i iii , , , ,
nnnLLLL p r s c oosn gggg LL1 L i iiCCCC
) m a ei rrii eeee OO0O EEEE 2 o m F S DDHS RRRR BBEE 468DDDD C m f u )) )))) )))) )))) )))))))
o h S 1 2 1234 1234 1234 1234567 t 2 i e W g d a e P z ( y l 2 a
- n 3
A s G. t _ 5 n _ 1 e d _ e i l c t n t b c n o n a A e i e l i t n d _ f s a o i o n k k r t i c a n n e n y t c y r f r B a B a p o d e l b a u n A a s i m t o t c e n _ m f d d i c Si n u o o o m c s A Ct a t R R A s n R a l S n d A o 1 z o o rl rl o r i $ l i o i ea ea r e d t ar C _ t ww oa ww oa l t a d e c e s u t u ns f ip Pr P r o W a j o es o r d d r o E i d e _ c oh t h t d l r i r s _ s rt biuit n l s d u cp s _ le ei o o i o p ce o f ZW Af. W C C l f R S AD l n o 1 2 3 4 / 8 I I 5 _ ui . . . _ . . . _ n. t 4 4 4 4 4 4 S 6 6
' c . . . . . . .
- e 5 5 5 5 5 5 5 5 5 l S 1 1 1 1 1 1 1 1 1 i
h Table 15.0,3-3 Summary of Comouter Codes and Methodolonies Used i i n Accicent Analyses i Computer Code Transient Numbars ** Analyzed with 1 or Hethodolo0V that Comcuter Code or Methodoloay i LOFTRAN 15.1.2*. 15. 1.3*, 15.1.4, 15.1.5" W.G.r.iW8
,e ,, u , 15,.2.6, , , e , , 15.2.7, 15.2.8* .
. . . _ . _ , ..._.., ..... , s5. M.
- 10. 4. 2 ;it, 15. 4. 4, 15. 5,1, 15. 6.1 r . - , , u,
., e ,o 6 15.2.1, 10.^,. , s.*.., .;. 2. 3 A
. . m. ,.. e, . . . ., ....,..
5 NOTRUMP 15.6./ a-c s SATAN-VI 15.6.# d g , E WREFLOOD 15.6.# d g 5 POWLOCTA 15.6./ d g 5 LOCTA-IV 15.6.p* l s ' L LOTIC 15.6.7 d g 5 LOCBART 15.6.4 d g BASH 15.6.)dg 5
- DAD 15.6.# d g
' ITDP 15.1.2*, 15.1.3*, 15.2.^a" . . . -
,e , e
..._,...,....-..z.4 l 15.5.1, 15.6.1 l- dB-1 15.1.2*, 15.1.3*,
1;.:.:' - ..
. . . . . , .. . . , ... .., ... . .c.
-1".'.:*, 15.5.1, 15.6.1 W-3 15.1.4, 15.1.5*, M '
To sed. -Ceo- hti,od-e) - poc3e,
]
** Transients are numbered according to tne cases listea in Table 15.0.3-2. ;
- All cases of- this transient used the comot;ter code.
N wt ,-* avvt-w+,m- u.m-- * - - -,-w e e V u' ttw.e'--- +warr-we -et--m-,es-w+--wgw w w --n e v .-*w=,- h~aerW'-+&tyv ' u a er - ,e'ee * * - --*-'-etv'we' e+ W ef N Ne&W
RETRAN-02 15.2.3, 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3d VIPRE-01 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3 SCD 15.3.1, 15.3.2, 15.4.1, 15.4.2, 15.4.3 BWCMV 15.3.1, '5.3.2,
. 15.3.3, 15.4.1, 15.4.2, 15.4.3 W-3S 15.3.3, 15.4.1 PDQ-XD 15.3.3, 15.4.1, 15.4.2 NODE-P 15.3.3, 15.4.1, 15.4.2, 15.4.3, 15.4.6 BWCMV 15.3.1, 15.3.2, 15.3.3, 15.4.1, 15.4.2, 15.4.3 i
1 i-l l 1-l ',
f : l t ; ;; l! l.j .
; ,l<j ! lI i i ! f g
e
-s ru et t a
- s. r )
t wep
- l 0 4
0 7 0 4 h t 4 0 4 0 4 0 8 0 3 0 8 0 4 0 4 3 M
?
S s em ( 4 4 4 4 4 4 4 " ? - /
.e l R
1 r e t e s Wu 3
-m l
)
t 2 9 4 8 2 9 2 9 2 9 2 9 4 8 4; 4 35 827 s-a4 i
? "
ro f 0 4 0 0 0 0 4 4 2 11 ! ! P eV ( 1 1 1 1 1 3 zH-a ! !
. )
- s. a .
0 0 0 0 0 0 0 0 0 0e 0 0 0 re: is 5 s e PP e r p ( 5 2 2 5 2 2 5 2 2 5 2 2 5 2 2 5 2 2 5 2 2 5 2 2 5 2 2 23 IM 3 2 2 5 2 2 5 2 2 s y l a n l A t e.
)
F I 0 7 1 0. 1 0 1 0. I
- 0. / 1 8.1 8r
/ 10 19%
I 0 1 0 n e
- 9 5 9 9 9 9 5 5 b1 " J
)
5 l e a VI ( 5 5 5 5 5 5 5 5 55' / 55 5 5' i f c o 0 0 0 OtM '1 00 c 1 1 1 1 1 1 0 0 0 A ) 0 0 o 9 9 1 r wm o p e e e e e e 0, 0, 0, o,N 5S t , 9 S. l - , , t t t t t t e o t g o o o o o o 7 7 7 0t 33 t g f Hf ( N N N N N N / 7 7 2 e 79 m 9 "2 a , l 3 3 3 43 33 I .Ii . P . ( s e 4 t l
- e a t s
3 m i i ) t f 1
. a t 5 p t 5 0 5 5 5 5 i. O 0 5- gh A ',
0 -s iSt w 2 2 2 2 2 - 2 2 ? 2 5 P a nS u H I NO ( 4 3 4 3 4 3 4 3 4 3 y.S M^ E' 8 33 -
?
^
3' 1 t e u s y f p r r m m m m m . 3kup: l b n e u u u u u u 3 3 : - _ m m c- : a i l m m m m - . T p i i i i i i e e e r- ' f p n n x n x n t t t o o i i a i a i o u o T- .i- i L D H M H H H H N f l N C N o. M M ; y r - - a m ) m c u c S r / o g t y / -
- i. t k 3 3 3 3 2 2 2 h . '
41% M.. ri / 4 4 0 4 0 4 "M M es e e e - e e n n % .. t 0 0 0 0. 0. 0 t t t ) MD 4 e t i o l o l u
- f. t. 0
( t f f
# a e
r ru ) ut t a f
- i. r * -
A A A A A A A A/ A 7 a re / lep m 4 1 Al f N 4 1 M i N l i 4 1 I P4 N ' H n< m c e g l Ht ( a i c t i n a b a b c d a b a b ; 1 r o 2 2 3 3 3 J 4 S 5 3 3 3 % c i . . . b u t l t 1 1 1 1 1 1 1 I 1 2 2 2 '- c . . . . . . . . . . . . S A. e 5 5 5 5 5 5 5 5 l S 5 1 5 1 5 1 5 1 1 1 1 1 1 1 1 1 el
ii !ll I! ! L;i:li((I i,. }l!j' ,, l ;i ,f! : i!! ,! tii ;ji e r ru - et 3 # t a i - ar ) - e. we dp " f 5 5 5 5 0 0 0 oN t o o / 4 4 4 4 4 4 4 A o A A A 9 em ee { 4 4 4 4 4 4 4 4 sL N 4 i t 4A 9 8
/
fT 1 1 t t 9J 2 1 r e
- t e C am )
l u l 3 5 5 S 5 o2 o2 o2 q - m 4, a L 1 / 1 3sr s9 s 9 ur o rW l / t ro f 1 A 2 2 2p 1 zV ( 1 1 I 1 1 1 qIf qD1 s91 g)Mu ; 0 A 4 N A 4 t P t t f 1 i -
. )
s a re s i s 0 8 0 8 0 8 0 8 0 5 0 so& e 0 0 0, 0 0 0 0 t o 2 z s zr p 5 5 5 5 5 5 5 e PP 2 2 2 2 2 2 2 '2 2 2 2 2 3 ( 2 2 2 2 2 2 z. 32 2 2 2 2 2 2 2 2 s y _ l a - n l A e 3 3 3 3 s I 1
- s g) . .
- 8. T a. 0 T a. T s s
gT8.Mis t n s vF e 4 4 4 4 0 - 7 t 0 0 0 S
)
f i d c e VI a* ( 9 5 5 9 9 5 99N59m94 55 493 I55 5s 5 eW7 t i e 5 9 994oe9 555.6 . o c 0 0 0 0 0G oG 0 0 6 oG%04o, 2 ** 00 A 0 0 0 m 0 0 0 1 1 o 0 % 2 ) 0, 0, 07 e r o wm 0, 0, 40 G o u, s,0
; t 57 ,9,9, E
u, 0, e e c 0 5 P g a f s So Cl RF h p /
; /
3
/
/
3
/
/
3 7 7 1. H& a O" 71 313s233m 7 a " 'l 3 2" 3 0 3 3 t t i o t i t os s,M 3 1 3 8 h3 5, 7 ( r e 4 t l
- e n
a t v E 3 i i u ) a tS p t 0 0 0 01 S1 5 1 i e 5' rI n
%5 3 5 r 0 b 5 9
a i St nS u N 5 6 5 5 5 1 4M 4 f 9 t 2 2 ss40' MR + 3 33 334 9334 6 6 6 5 1 l' I NO ( 3 3 3 3h % . 4 3 4 3 a1 33 . 3 _ e t u y r r e r r o l b p n r e m u u t u m u m u
/. .
/a Y* ;. q5. -
/o A a i l m m m m "f L t oS . E1 c l p i i i i e t- *; m .
h f e p x x x x9 ehR f 9 9 o =- t o o m A A A C y D N N N N0i - 0 ob i t N ;" t i 4 1 4 f f 4 9. 0 JW o r - a m ) m u c S r c
/
o u 4 1 4 4 1 1 4 i s _ t y / . . 1 l i at k e t a 4 8 2 _ ri / t lief l(ho6 JM 1 A 1 es o o d n n A 4 A4 A4 A4 NJN8 FJ 4 4 ec- n A A A 4 lr6 s o1S 4 i i _ I 1 1 t 3 oe I ( 1 I t f DJ MD i t _ e r ru _ ot ) t a F 4 ar a re / / / / / Ar nf Af e A A A A A ep m dm c oe p e *
- e J
}p Nf Np t l
o 4 1 @ 4 i 4 f 4 1 1 4 s f MT ( n a b 4 / - a b c d b o 6 / 8 8 3 R i . . . . I 2 I 2 m J 3 3 3 3 2 2 2 2 A c t 3 3 3 4 4 4 4 4 4 3 e 5 F S 5 1 5 1 S 1 5 1 5 1 5 1 5 1 5 1 5 1 4 5 1 5 1 5 1 5 1 s i
\
s << ' , ] l : :! 1j j j e ;! i ;i ;
i Table 15.0.3-4 (Page 5 of 5) Summary of Irout Parameters for Accident Analyses Notes
- 1. The results provided corresco.1a to an assumed initial reactor coolant f'ow of 393,600 gpm. Technical Specifications require a measured flow of 388.880 gpm. The dif ference between assumed and required flow has been justified and approved based upon the ONBR margin available and based upon an evaluation of DNBR sensitivity to flow submitted with the McGuire Unit 2 Cycle 2 and Unit 1 Cycle 3 reload design transmittals.
- 2. The reactivity coefficient assumed is a function of moderator density as described in sections 15.1.4.2 and 15.1.5.2 (Figure 15.1.4-1).
- 3. Refer to Figure 15.1.5-1.
H.\ 4 RefertoSection15.J.t.2.
- 5. Consistent with the lower limit shown on Figure 15.0.4-1.
- 6. - ;cdition to thc 'ul' couer (3C5 Wt) a5u, ,env i s. u i o nuc+es -cre run 44P-9aftial scwur l eve i > vf 60k (1055 i-Twi.) and 10% (343 ,MWi.h
- 7. !^ addition to i.iie ivil pu-ec (030.1T) casu , mens i c i v i ty >cudie; ucre run for pectial pvwer iuveia vi 00%- (G76. 0"F) and lff# ('.00. 2 I).
- 3. hr-eedi tion to av iuli svwer (1092 i9) case- ;enaitivity stucies -cre run fer oe-tial powec 'eveis vi 60% (G40 f f) snc- 10% (:44 f t').-
- 9. 4m addicion T.o T.ne fuii sv-er ( MO O ma e m it4"ity ;tudie5 -er; run fon-rart i a l s voer ieveis vi 5 A--( %0 " anc 10% (241""'
- 10. Refer to Section 15.4.8.2.2.
- 11. The mocerator density and Doopler effects on reactivity during LOCA transients are accounteo for in the evaluation models as descricea in Section 15.6.5.2 ana the associatea references.
- 12. The hyaraulic analysis was performea using 102*. of tne Engineered Safe-guarcs Design Rating. The core thermal transient analysis is performed using 122% of the NSSS licensed cower.
- 13. The 3 puno flow expected for 4 oumo flow of 388.800 gpm is 282.300 gpm, but toe actual LOFTRAN analysis used 284,460 gpm. The results of the analysis are not sensitive to sligntly nigner initial ficw.
H T h e- ~ o sT, posi b t WTC. ( more n9Q mQ @ w~ w w.A% m, a, e DOI Q h 4 0 >= k,,, ,
-i+rtt
De.Ich bc3t. , TABLE 15.0.3 5 Nominal values of Dertinent Plant Parameters Utilized In The ACCicent Analyses Thermal outou of NSSS (MWt) See Table 15.0.3-1 Core inlet temper ure ('F) 559.2 Vessel average tempe ture (*F) 588.6* I Reactor Coolant System p sure (psia) 2250 i Steam flow from NSSS (lb/hr) 15,140,000 Steam pressure at steam generato o let (psia) 1000 l Maximum steam moisture content ) 0.25 Assumed feedwater temperatur at steam 440 generator inlet ('F) i Average core heat flux tu/hr-ftz) 197,200 Control rod drop ti . (sec) 3. 3
' A 1.5'F allo nce for steam generator fouling facto is acdea to this numoer to rive at the value given in Table 15.0.3-l l
l 1
- d::
_ _ _ . . _ _ _ _ - _ _ . _ _ _ _ + _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ - _ _ _ _ .
- . - - .-- ._. .- . . - = _ . _ . - - - .- _ . _ _ . _ _ .__ _-. _ - . _
L & s N. Table 15.0.3-6
\ Values of Reactor Coolant F1 Utilized in The Accident Malyses s
l
'N
'N. With Without Improved Improved
-\ Thermal Thermal
/ Design Procedure Design Procedure
'N Reactor coolant low per loop (gpm) 98,400 94,250 Total tctor Coolant flow (10elb/hr) 146.1 l'40.0 s {
N
's,N l
l e
& w w pr'
Table 15.0.5-1 Rod Droo T{mes Used in FSAR Analyses Drop Time to Dashpot FSAR Section (sec) 15.1.2 3.30 15.1.3 3.30 15.1.4 Note 1 15.1.5 Note 1 15.2.3 3.30 . 15.2.6 3.30 15.2.7 3.30 15.2.8 Note 1 15.3.1 2r70* 2.2 15.3.2 .2,-70e 2.2 15,3.3 L-77 2 2. 15.4.1 3r30 2- . 2 15.4.2 3,-3r 2 & 15.4.3 1-30 2 . 2. 15.4.4 3.30 15.4.6 Note 1 15.4.7 Note 2 15.4.8 3.30 15.5.1 . 3.30 15.6.1 3.30 - 15.6.2 Note 2 15.6.3 . Note 1 15.6.5 3.30 15.7 (all sections) Note 2
" Reduced crep-t4me-assc;.e3 cen. iuent with--Ic-w cei . ' low cond4%n-at '4ma of reacter te+pr Note 1 Results of transient are not sensitive to rod drop time.
Note 2 -Refuttr trip was not necessary to analyze transient. R ea.c.ko v 10cc upeate-l t I _ . . . _ . _ __ _ _ . - _. __ L
Table 15.0.6-1 Trio Points and Time Delays To Trio Assumed In Accident Analyses ' l l Limiting Trip i Trip Point Assumed Time Delay: l Function in Analys Q (Seconds) Power range high neutron flux, high setting M ** 0. 5 Power range high neutron Ill.1 flux, low setting Jtg % 0. 5 High neutron flux, P-8 85% 0.5 2 Overtemperature ai Variable see 10.0* < Figure 15.0.3-1 j Overpower AT Variable see 10.0" Figure 15.0.3-1 High preso aizer pressure 2410 psig 2. 0 Low pressurizer pressure 1835 psig 2. 0 Low reactor coolant flow 86.5 (from loop flow detectors) g% loop flow 1.0 Undervoltage trip 68% nominal 1. 5 Turbine trip Not applicable 2. 0 i ! Low-low steam generator ** 4.0 i level l High steam generator level 87.4% of narrow range 2.0 feecwater pump trip, . level span feecwater isolation, and turbine trip
- Total time delay (including RTD time response, and trip circuit, channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to fall.
**The numerical setpoint assumed for this trip function varios depending on the accident being analyzed. The values used are given in the descriptions of the various accidents.
1986 Update
Ddek
\
9e / xx Table 15.0.7 1 /
/
'N Determination of Maximum Overoower Trio Point-Power
\,
Rance Neutron Flux Channel - Basca On Nominal Setooint
\ Considerino Inherent Instrumentation Errors N
(1) Calorimetricherrors in the measurement of secondary sy em thermal cower: Estimated Effect on Measurement Accurac Thermal Power Deter-Variable \ of Variable (% err ri mination (% errer) Feecwater flow \ 21.25 !1.25 Feedwater pressure N (Small correction \ on entnalpy) \ -10.5 Steam oressure \, (Small correction \ on enthalpy) 2 Feecwater temperature !0. 5 Total non-flow effects 13.0 \ 0.3 Assumed calorimetric error (% of rateo pow ' (2) Axial cower distributic ef fects on total ion 'etiamoer current
/
E,timateo error (% q( rated power) NN 3 L Assumed error (% o rated power) \ 5 (3) Instrumentatic channel drif t and setcoint r eprocuci i'tity Estimatea error (% of rated power) Ny 1 Assumea er/or (% of rated power) \ 2
/ \
~o tal assu-Ma errors in setcoint (5 af ratec cower) 'y (1) y(2) + (3) 29
\
/
Nomin3i satooint (% of r3ted power) 109 ' Max um overpower trip setooint assuming all individual \ er'rors are simultaneously in the most aaverse airection \ d of rated cewer) 118 \x f , N ' [. l 1986 Upcate
Table 15.0.9-1 locine and Noble Gas Inventerv im Deactor Core and Fuel Dod Gaos' Core Inventory Fraction of Inventory Gap inventory Nuclide (Curies) in Gao** (Curies) 1-131 r.9to? ??OEOS . 10 f.i t' <W I-132 i. 3 1MiE08 .10 t3 ME07 1-133 f.9 #4E08 .10 19 M E07 1-134 LI M E08 .10 z.t R$E07 1-135 f. ? M E08 .10 f.8 2ME07 Xe-131m t.3 1. 06 .10 t.3 ME05 Xe-133m s. 4 6 LEO 6 .10 s.& sdE05 (e-133 t. 9 1E08 . 10 f. i ,2ME07 Xe-135m 3.5 :;. ' E07 .10 2. 5 MA06 e-135 ye.g37 Q' Xe-138 I,,D . 9 L E08 i
.lO ?.10 g,y ,, f .if42.4E07 M07 i.4 )..$E08 .10 Kr-93m i.I haE07 .10 f.4 OME06 Kr-85m z.s M07 .10 r.5 M E06 Kr-85 49: eE05 .30 4.5 ME05 Kr-67 47 51 E07 .10 4 7 S-4E06 Kr-88 6.5'.yE07 .10 6.f L4f06 Kr-89 f.-2 lE07 .10 f.2. k4E06
' 3aseo on an eouilibrum cycle core at end of life. *he w.:r :;i:r core coerates at a cower levei of M Pat inc an average cycle ournuo ;*
WI 10,{0 WD/MTU.
" NRC assumotion in Regulatory Guide 1.35 007 .-coate
Table 15.0.9-2 Reactor Coolant Scecif4c Activities f0r !0 dine and Nobt e Gas Isotoces Nuclide Soecific Activitv a (uci/c) 1-131 54 o.(,(e I-132 $
- o.24 l*133 % l* l I-134 .94 J.14 1-135 M o.S8 Xe-131m 1. 9 Xe-133m 3.1 Xe-133 281.0 Xe-135m 0.7 Xe-135 6.3 Xe-138 0.7 Kr-83m 0.0 Kr-85m 2.1 Kr-85 8.8 Kr-87 1.2 Kr-88 3.7 Kr-89 0.0
- Reactor coolant concentrations at eauil(brium assuming -1 : :: : '_
t
- Cal % ficad joes ,T d i,4 e d k k ,
uc7 u = w.
hem gg w +A Table 15.0.12-1 Offsite Doses (Rem) [ 2-br Dose at 2500 t t. 30 day Dose at t. x Exclusion Area Boundary low Popui Zone FSAR , Section khole Thyroid Whole B y T hyroid Accident 15.1.5 2.5E-03 3.3E-01 E-04 9.2E-02 Main Steam Eine Break 'N Loss of External Load N 15.2.2 2.4E-03 1.9E-01 8. 5E- 04 8.lE-02 Rod Ejection Accident 5.4.8 6.6E-02 8. 1.lE-02 1. 7 Steam Generator Tube Rupture 15.6.3 2.8E-01 4.1 2.5E-02 4.2E-01 Eoss of Coolant Accident with ECCS Eeakage without ECCS Leakage 15.6.4 h[ ' 4[21\ 2.2E*02 2.0E*02 8.0E-01 7.9E+01 4.8E+01 4.0E*01 Waste Gas Decay Tank Rupture 15 7. 5.0E-01 - 4.4E-02 - Liquid Storage Tank Rupture .2 ". 9 4.9 1.3E-01 4.4E-01 fuel flandling Accident 15.7.1 1.1 5.lE+01 9.6E-02 4.6 Inside Containment fuel Handling Accident in The fuel Building 15.7.4 fuel Assembly Drop 1.1 5.1E+01 9.6E-0? 4.6 l Weir Gate Drop cGuire) 3.5 2.0E+02 - 2.3E-02 5.5E-02 Weir Gate Dro (Oconee) Tornado Mi ile impact ISB 9.5E-01 2.7E+02 - On Spent uel Pool 2.3E-03 2.9 2.7E-04 2.6E-01 LOC dith Coincident ISB er Containment
'ressure Relief 1988 Update i
i!i i!![i i{,if I !f .)I ,L [ et f ii!!!t;{l! ! IIIiI t' f ) ) ) ) f 0 0 ) 0 i ) 0 ) ) ) ) o 0 0 0 0 0 0 0 0 8 90 0 0 r 7 . 0 4 0 5 0 0ey 10 0 90 0 0 80 0
, nh . 3 7. 3 3 2. 3 7. 3 3 7. 3 7. 3 9. 7 9. 3 0( 2( 8( 0( 3( 3( 0( 0( 6( 6(
9oT 2Z tn ao .i et say old DuopB 2 ) yo 2 3) 8 ) ) ape 0) 20 7) 2) ) 0 2) 20 ) 0 d l 05 0 15 05 55 5 55 5 70 7 . wo 0. 2 0. 2 5
- 0. 2 0. 2 2. 2 2. 2 5
- 0. 2 05 2
5 6 5. 2 5 0oh 0( 0( 0( 3LW 0{ 0( 0( 0( 0( 0( 0( s ydi ) ) ) )
)
e ) 0 ) 1 ) 0 ) 0 ) 0 c .r 0 0 3 n .ao 0 0 0 40 5 0 0 0 0 f e tdr fny 0 0 0 0 0 0 0 0 .5 0 o u uh 2. 3 2. 3 3. 3 0. 3 03 03 0. 3 0. 3 97 93 c 7( 5( 0( 1( 1( 2( 2( 2( 2( 1 e 0oT 1( ( s 0B n 5 1 o 2a C e tr 7 l aA a v 0 t end C r soo 1 e oiB
- 5) 6 ) ) )
m Ds 6 0 n ue 1) 80 4) 1) ) 0 ) 0 ) e o rll 05 0 15 05 5 65 6 0 5 l hco '05 8 85 5 b i r
- xh 0. 2 2 0. 2 0. 2 2 2 3. 2
- 3. 2 2. 6 2. 2 a v 2EW 0( 0( 0( 0( 1( 1( O( 0( 4( 4(
T n E n 8 o 5 2 3
- 3. '
i 2 3 4 4 Rt 1 Ac 5 5 5 5 Se 5 1 1 FS 1 1 1 k e d l ae n a e ae e en i o n wn n ri d L i ain i Bdn o dn rdo dn oo doi te oo~ te oo te I l te eIi nk a Ii nk hit nk Ii nk n t ei t n t ei t a ei . t ei . ica dp n r ca dp icr dp ca dp Ler i S e e rer i S Wet iS ner i S _ pt c r t o pt c pn c opc c mSn ce .r x tSn ce dS e ce iSn ce a e An ue E o e An o c An t . e An i .ck Rhc i .Rhn i chc i t ehc ed . co ecn ed n tcn ea ro ni op f o deo cn ro eeC ed ro jeo ro e Seo ETC PI d TC PI CS eTC PI lT g PI - i n s k
.c i. . . s c . .' n . . d' . .
c aa b c o oa b ia b oa b A M L L S R il ; i ' ,: lj ,< ..I i ll';ji]jl;ti
t ) ) ) f d ) ) 0 0 0 i ) 0 ) ) 0 ) 0 o 0 0 00 9 0 0 0 0 _ 0 r 6 0 0 0ey 10 90 0 0 0 10 0 0 0
, nh 3 3 2. 3 0. 3 0. 3 3 23 73 13 _
0( 9( 2( 0( 1( 1( 3( 2( 3( 9oT _ 2Z tn aoi et say _ old _ Duo pB yo ) ) ) ) ) ape 51 80 9) 5) 0 ) 0 0 0 d l 15 6 35 35 4 . 75 wo 0. 2 0. 2 5
- 0. 2 0. 2 15 2 3. 2 4. 2 5 45 2
45
. 2 0oh 3LW 0( 0( 0( 0( 0( 0( 2( 2( 2(
s ) ) e yd ) 0
)
) ) 0
)
) 0
}
0 0
)
c ri 3 n .ao 0 0 0 5 0 0 0 0 0 0 f e tdr fny 0 00 0 0 0 90 00 00 00 o u q uh 9. 3 13 43 3. 3 2. 3 3 53 1( 43 1( 43 1( 1( 2( 0( 6( 1( 2 e 0oT 1( ( s 0B n 5 1 o 2a C e tr 7 l aA y 0 a t end 5 n soo 1 e oiB ) ) ) ) m Ds ) - e n ue ) 0 35
)
45
)
5 0 55
) 0 0 0
- l o rll 75 1 7 5 4 35 5 5 5 r hco 3. 2 b i - xh 2 7. 2 2 3. 2 2 22 22 22 a v 2EW 0( 0( 0( 0( 0( 0( 1( 1( 1( T n E n 5 o 2 3 i Rt 6 6 6 Ac 5
. Se 5 5 FS 1 1 1
_ t _ n e e' e n d r e) e n e i c tA n i n i i d' i d c fV dn o dn o~ A a . k e oo te I oo te I S o aed Ii nk r Ii nk t C Sn erit t ei t o .t ei t n C C( rusnca dp n t ca dp n a S E C Bsteer i S e aeer iS e l C Es c o s u m pt c r rrpt r Ce te r eeOnSn ce r euSn ce r' o Eg ug mu nr i . e An ue nt e An ue C a oa po t_ iPyahc i ck e ph c i ck hk hk gh n L rtcn ed ni Gucn ed ni f eta ta e Saneo ro op Reo ro op otie ie 04 d lCdoTC PI CS m TC PI CS iWl Wl 32 i lRnC ae ss c a u . . . eb . . . sf . . c c mnof a b c tua b 2 ofa b A SiBo ST LO
..L i L l
,: ;Ii
Table 15.0.7 (3 of 3) -
- f I Environmental Consequences ;
- . 2-hr Dose at 2500 ft. 30 day Dose at 23,000 ft. I FSAR Exclusion Area Boundary Low Population Zone ,
l ! Whole Body Thyroid Accident Section Whole Body Thyroid' 1~ I . Loss of Coolant Accident - 15.6.5 [ , . Control' Room , a'. With'ECCS. 1,2 16. leakage (2.5) (30.0) [ i
- b. Without ECCS: 1.1 .13. '
! leakage (2.5) (30.0) ;
- 25. ;
- c. 50 gpm ECCS after 1.2 -
L i 24 hours (no VA)- (2.5) (30.0) i t ' Waste Gas Decay 15.7.1 0.41 N/A N/A N/A , , Tank. Failure (0.5) 8 i Liquid Storage 15.7.2 . Tank Failure
- a. Tech Spec, Iodine 2.0 1.2 0.18 0.11 [
Concentration (2.5) (30.0) (2.5) (30.0) : t I ! Fuel Handling Accident 15.7.4
- a. Dropped Fuel. Assembly . 0.80 27. 0.071 2.40 l
- Inside Containment (6.0) (75.0) (6.0) (75.0) I i
i j b. Dropped Fuel Assembly 0.80 9.1 0.071 0.81 j Inside'SFP Building (6.0) (75.0) (6.0) (75.0) l l 0.081 1.7
- c. Dropped Weir Gate 0.91 19.
Inside SFP Building (6.0) (75.0) (6.0) (75.0) !
)
- d. Tornado Generated 0.77 190 3
j Missile Inside SFP (25.0) (300.0) i
- l. Building L
r i V l. i _ ,- . . , - - , . _ _ . _ . , . __ ., , __.,. . ~ _ _ . -
y LC .
% s s s s s s s s
{j s s s s s s
#6 .- -- N'- ' s s s
' .,,,* s s
64 17731' 2000 , OPAT SIA s s PSIA s , ,\' ,' 52 s s s sN s s s \ S 60 g s s s \ s
\ 2250 y 58 ,
s s s s s PSIA [%\s s s g N' T
< 56 s
s s s s
\
~
s s s d s s s a :.4 s s s ! s s s s s s i 52 s s s s s s g s s s 50 s s s s s s s s s 48 s s 46 Steam Generator s Safety '/alves s s 44 Ocen , s ( s ( s s s s s
, i 42 535 590 595 600 605 510 f.- 620 625
- !,es ,
avg - ~ , l
- - - - CTAT , CORE LIMITS I
l' L l ILL"STRATIO: OF CVERTEMPERATURE t AND OVERP0b'ER iT PROTECTION in ' Saupy McGUIRE NUCLEAR STATION ' T1;ure 15.0.3-1
;904 Update l
l 1 _ _ . _ _ _ _ _ _ . _ _ _ .
- . _ . . , . _ _ . _ . _ _ . ~ . . _ . _ . . _ _ _ _ _ . _ _ _ _ _ _ . _ . - . . _ _ . _ _ . - - . _ _ _ _ _ _ . _ _ _ . _ . . . .
i
+ \ \ \
\ \ \
\ \ \
\ \ \ \
\ \ \
_ _____\\ m_g \ % Low Pressure Rescuor " Dip 9 W \
\
\
\
g g
,c,,7 '
. =
\ \ \ \ s Pressues i
\ \ .
' Dip
\ \
ga [% i .
- ', ', Opossang '
. \ \
Pressws g ,
~ \ s % %
>-- % % g\
< s \
t-
* \ \ \
.d wd \ \ \
O \ \ \ \
\ \ \ \
\ \ \ \
\ \ \ \
\ \ \ \
\ \ \
\ \ \
\ \ \
$ \
\
\ \ \
*\ \ \
\ \ \
Steam Generstar ., s Safecy Wlves Open 4 4
\ \ \
\ \
\
\ \ \ '
T
. avg ( II OTAT CORE 1.IMIT3
)
ILLUSTMTIO:1 CT CVERTOTERATURE
/d;D 0VERPOWER ;; ??.0TECTION i McGUIRE NUCLEAR STATION-w Tigure L5.0.3-1
.?c4'U; dace p-.
._ . _. _ . _ _ ___ _1_ _ -_. _ . _ _ _ . . _ .
l 1.2 1.1 -
- t O / !h 1.0 -
FULLYtNSERTED
/
p l 0.9 - B,C SCR AM CURVE - I k \
- l
\9 l~~$n 0.8 - l C I OASHPOT wo ,
I- ~ 0.7 - 5_ e l (53.6 - l l ME j3 3.5 - l 9 "- I
, y 0.4 -
l I l m 0.2 - u l 3
< I g 0.2 -
l
* \
0.1 - i i i i i 0 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.3 1.0 1.1 12 13 14 15 TIME AFTER DROP BEGINS/ DROP TIME TO TCP CF CASHPOT + RCCA POSI!!CN VERSL'S THIE TO DASHPOT
'un nen- ?1cGUIRE NUCLEAR STATION Y Figure 15.0.5-lc4 j
'M'. L'e m e I
,_..I_.
__ . _ _ _____m.__ . . . ~__m .m. , _ . _ . _ . . - ~ _ . . -_-_..m-.--__.- -- e I a e i 2 T ~a.es lM _ f* 2 - E 5.a
** M o == o L
b 2 m
- n.
C
===9 e O o
_ y-
- W -
*== m. A >=
0 O ac A QO co ar.
~
o bk N m r
~ C --
~"" e f x La,a 8"
- an A O g
o a l-oc 4 w m ~ o g" N_
- 9 o
, N.
O E 1 l C
- l. l l l- } ! ! -! ! !
. , _ o
~ ,
o, - m. .,. 3.. .. .o. . m. o
~.
o o
. _ _ _ . e e -- o- e o o (10dHSVO iO dO10130NY1S10/03ddOHO 3DNY1S10)
N011150d V008 03Z11VHWON RCCA
- ... y:. .. ,.PO.SITION VERSUS vn5.nr. 0, W$G88 M l ricure 1 o. 0.2..1 b . _ _ - _ _. _ . . . . _ _ -. - __. . . _ _ _ _ _ _ _ . _ . . _ _ _ . . _ . _ ..____i_____. _ . . _ _ . . _ , _ _ .
\
s I.0 0.3 - 0.8 - g 0.7 - I
$ 0.8 -
E G g 0.5 - a it 0 0.4 - E
,)
i R - l 0.3 l 0.2 - 1 0.1 - l
' I I 0 i 0 0*2 0.4 0.6 0,s g,o ;
i 400 PC51T'CN (RACTICM INSEtTED) l' N0iWALI::ED POD WORTH VERSUS I PERCENT r.iSERTED- ! 4"*"! McGUIRE NUCLEAR STATION Y Figure 15.0.5-2ca . 1984 Update
NORMALIZED ROD WORTH VERSUS PERCENT INCERTED 1 0.9 - 0.8 - l l 1 0.7 - 0.6 - t E 0.5 -
- e S i l
0.4 - !
=
l ! 0.3 - 0.2 - l l 1 01 - 0 O 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 ROD POSITION (FRAOTION INSERTED) NOR!"Jd.IZED RCD WORTH VEISUS PERCCIT INSERTED Mk ovu m e' McGUIRE NUCLEAR STATION W' figure 13.0.5-2h
. . . _ _ _ . _ _ . - _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ . _ _ . _ - . _ _ . _ _ .-~- _ _ _
t i
+
r i
)
L 1.0 i I
=
2 l l j 0.8 -
.l ;
l i
- -u 0.6 -
i d l
< l u '
y 0.4 -
=-
w l 2 I a I j 0.2 - 3 l QASHPOT 1 k I O-
' ' I ' '
' I 1 I l-0 3.1 .0.2 0.3 0.4 0.5 0.6 0.7 0.8 3.9 1.0 '1.1 1.2 1.3 1.t NORMAll:ED ROD CROP TIME TIME AFTER ORCP BEGINS TIME TO TOP OF DASHPOT 5
- 0M!ALICED RCCA 3AW REACT;-
'lITY WORTH VERSUS ::Ottg;;;;
i L OROP !;:!E
..u rconiA 'icGUIRE NUCLEAR STATION 71:ute
-=
15.0.5-3m
,, g I
..-...;... .~. a.. - - ..-- - .-.
NORMALIZED RCCA BANK REACTIVITY WORTH VERSUS NORMAll2ED DROP TIME 1 , l 0.9 - 4 0.8 - { 4 0.7 - A . 0.6 e - 3 l d O - { h.5 a 5 g0.4 - 4 0.3 - 4
)
0.2 - ./
/.
/
O.1 L-
, DASHPOT d .
0 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 NORMAllZED ROD DROP TIME
- 10RMALCC RCCA 3A:E EEACC--
VITT *' ORTH VERSUS ::0DU.LZ~J 2 ROP C';
..b
'unt an NicGUlRE NUCLEAR STATION
___li;gra 4 9.@ .3-%
15.1. 5.3 Envira-mental Cansecuences t ac;rc3rv .s-
' e :sta.atec acci:ents .nvolviac release
- iteam # am t*e tam ao not esult , a release of rioloactivity unless tnere '
5 .eanaae ' am
*".e . RC5 :: ne seconcary system in the steam generators. c:nservative 3nalysis f tne ootential offsite cases resulting frem tnis acc10cnt is cre-1enteo c:nsicering ecu11ibrium coeration casea upon 1 percent efective fuet ina a 1 ;;m steam generator leak rate prior to the postulatea acticent.
hW\Mtt. if **n,mnO^^hd ,N. ; 2 .
.s . .... __ ;j
-- ,-.-m ,me,-smm ;; ,~,4.s..;; ;;7__,-mm- 7
__7-1 apr3 #4-,+t - ;
*; ; ; _77,wnmg,
*he f ollowing assumotions and oarameters are used to calculate the activity aelease and offsite cose for a postulated steam line break:
hema ys
- 1. Deior to the acc1 cent, an eauilibrium activity ofjsfission orooucts ex-ists in the crimary and seconaary systems caused by primary-to-seconaary
'eaKage in steam generators.
.. *he total pr1 mary-to-seconaary lean rate is 1.0 gpm, ith 0.25 gpm in eacn steam generator during the first 2 hours. After 2 hours, no leakage is assumed in tne faulted steam generator and 0.333 gpm cer generator in the non-faultpa steam generators
- 3. Offsite power 's lost 'anen :ne steam line creak occurs.
4 The steam release for the cefective steam generators terminates in 2 nours. The steam release from the non-f3ultea steam generators _c:n-tinues f or eignt hours.
- i. All neole gases -nicn leak :: tnc seconcarv siae are reieasea v'a tne steam release.
~5e taoine 03rtiti:n factor curing tne acc1 cent i s.Gef. 0 01 ther assumottens are listea in Table 15,1.5-2.
Basea on tne moael cescribed in Aopenaix 15A. the thyrola ana nole coov coses are calculateo at tne exclusion area counoary ano the low ocouiation ::ne. ~he
-esults are cresentea in Table 15.1.5-2. ~5e coses are witnin tne i"mits of
'.] CFR 100.
?3' J':28 15.1-17 3, g7
Three cases are considered for each accident to conservatively model the iodine that may be present in the primary and secondary sides. These cases are as follows: Case 1: 11ormal equilibrium Technical Specification iodine concentrations exist at the time of the accident (see Table 15.0.9-2). Case 2: There is a pre-existing iodine spike at the time the accident occurs. The reactor coolant concentrations are the maximum pernd tted for full power operation; i.e., 60 times the normal concentrations. Case 3: There is a coincider.t iodine spike at the time the accident occurs. The iodine concentrations are found by increasing the typical equilibrium appearance rate in the coolant by a factor of 500.
- - - - ~ - - - - - _ - _ - - _ - _ _ - - _ _ .
1 l Table 15.1.5-1 (Page 1 of 2) Parameters for Main 5 team Line Break Dose Analysis
- 1. Dati and assumptions used to estimate radioactive source from postulated accidents 3565
- a. Power level (MWt) 1 l.
b, Percent of fuel defectea I i l c. Steam generator tube leak rate (gpm) i 1
- d. Offsite power iM t a141abl0 1 i
- e. Reactar coolant activity Table 15.0./-2 l 9 i l 2. Data and assumotions used to estimate activity i released 01
! a. Todine cartition f actor during accicent 0.)f steam release t 181,000 l d. Initial steam release from defective steam generator (lb) (0-2 hr)
- e. Steam islease froia three nondefective l steam generators (lb) 752,000 (0-2 hr) 't,128,000 1 (2-8 hr)
- 3. Dispersion cata ,
Distance to exclusion area bcundary (m) 762 a. Distance to low population zone (m) 8850 b. l c. x/Q at exclusion area toundary (sec/m3) ' 0E-04 1
- d. x/Q at low pooulation :ene (sec/m3) 3.0E-05 I
i 1
,2, .
i
Table 15.1.5-1 (Page 2 of 2) Parameters for Main Steam Line Break Dose Analysis 4 Oose data
- a. Method of dose calculatior.s !rpendix 15A
- b. Dose conversion assumptions Appendix 15A
- c. Doses (Rem) See Ta\da. \S.O.'l-l N M usion area boundary Whole body 2.5E-03 Thyroid 3.3E-01 Low por J1ation zone Whole body 8.7E-04 Thyroid 9.2E-02
- 2. The primary to seconcary leak rate is evenly aistr .autea to the f our steam generators , i . e. , 0. 25 gpm oer generator.
- 3. The plant is operating with I cercent defective fuel.
4 Seconcary siae activity as a resuit of ortmary to seconaary !eakage nas reacnea an equilibrium wnca the acciaent occurs.
- 5. All noole gases wnich leak to the seconaary side are releasea.
- 6. The steam generator blowdown rate is 50 gom.
- 7. The turoine building steam leak rate is 1700 lbm/hr.
- 8. The condensate leak rate is 7510 lbm/hr.
- 9. The glana seal iocine partition factor ano the condenser air ejector icaine cartition factor are eacn 0.15.
- 10. Steam releases, wnich are the only significant releases to the environment, terminate in 8 hours.
- 11. The steam generator iodine cartition f actor during the accident i s jlPEf* CLOl
- 12. Water density at 180*F is used to convert condenser .olumetric flow to mass flow.
- 13. Other assumotions as detailea in Table 15.2.2-1.
l Based on the model in Aopenaix 15A, t.1e tnyrota and wnole ocay cases are calcelated at the exclusion area counaary ina tne low population ::ne. The results are cresentea in Table 15.2.2-1 a us are within the limits of 10 CFR 100. 15.2.2.4 Cenclusions Basea on results notainea for the turbine trio event (Section 15.2.3) and con - siderations aesc rced in Section 15.2.2.1, tne acolicaole acceptance criteria for a loss of external loaa event are met. ! 15.2.3 TURBINE TRIP
'.5. 2 3.1 Identif4 cation of Causes ana Acci:ent Descriction For a turoine trio event, the turoine stop valves close racialy (typically 0.1 sec. ) on lors of trio fluid pressure actuatea by one of a numoer of cossiole j turoine trio signals as descricea in Section 10.2.2.
l Uoon initiation of stoo valve closure, eteam f'ow to the turoine stcos aoruptly. 5ensors on tne stop valves actect tne turoine trio and initiate steam cumo. The loss of steam flow results in an almost immeaiate rise in seconaary system temoerature and pressure with a resultant primary system transient as cescribea in Section 15.2.2.1 for the loss of external loaa event. A more severe transient ! occurs for tne turoine trio event cue to the more raoid loss of steam flow causea by tne more racia valve closure. 15.2-4 7"' ?&
. . . - - -, - - - - - - -- . . _ - . . -- . - =_ - The automatic steam dump system would normally accomodate the excess steam generation. Reactor coolant temperatures and pressure do not significantly , increase if the steam dump system and pressurizer pressure control system are ! functioning properly. If the turbine condenser were not available, the excess ! steam generation would_be dumpea to the atmosphere and main feedwater flow - l would be lost. For this situation feedwater flow would be maintained by the l Auxiliacy Feedwater System to ensure adequate residual and decay heat removal capability. Should the steam dump system fail to operate, the steam generator I safety valves may lift to provide pressure control. See 15.2.2.1 for a further i discussion of the transient. A turbine trip event is the most limiting of loss of external load, loss of. l condenser vacuum, and other turbine trip events. As such, this event has been ; analyzed in detail. Results and discussion of the analysis are presented in i Section 15.2.3.2. 1 A turbine trip is classified -as an ANS Condition II event, a fault of moderate I frequency. See Section 15.0.1 for . discussion of Condition II events. l
- .5. 2. 3. 2 Analysis oi Effects and Consecuences )
l Metnod of Analysis j g In this analysis,lthe behavio of the unit is evaluated for a complete loss of i steam load from.1GO' percent f full power 9 :rily to show the adequacy of the ; cressure relieving devices .nd cis; t; d;;r.;n:tr:*. ccre pr;t;; tier, aergi-fts, i The reactor is not tripped untii conditions in the RCS result in a trip. No credit is taken for steam dump. Main feedwater flow is terminated at the time ) of turbine trip, with no credit taken for auxiliary feedwater to mitigate the l consequences of the transient. RETRAM. o 2. 4 puter program C ta}(ReferenceThe /).Theturbinetriptranientsareanflyzedbyemp program simulates _the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, ; steam generator, and steam generator safety valves. The program computes pertinent plant variables including temperatures, pressures, and power level. F 2:d de-* 4 s nelyzed mith the I~cyed NrW 5:ign sc;dro-n desc""ed k, in Reieic m 4 & Major assumptions are summarized below: fdl pr- obstaMoe
- 1. Initial operating conditions '4ti:1 h mod RCS terp;nt-,e.areassumedtoDeattheirb:::":t:rpn;r.;h;;.u -- b valuasdudncertair. ties, tn-- i n i ;. i o i uw i i mmr e re i Wem 4.. nc i m. L ::CR u ocai li in Eefiidusi $ *g
- 2. Moderator and Doppler Coefficients of Reactivity - the turbine trip is analyzed with"b th ::::= :-d :f-im~a reactivity f eecback. N: :r 'er
- t;d cases 2:te e = 1rge ps:itive cder:ter dca:itj c;;f'icient and tb :::: -ff.icim.d., The " % m sue + assumes aneca'" " --icr ,_:r :hWre coef ficient' and the leastfcewouK r5deratort^- onelyds negative Dopplers coefficientl k: 'i g e r; l' . 0. "-U . c.orrssponk*ngto b Q nng*,g la.mpuo h c4 c. ore, L.We,,
15.2-5 1986 Upcate
- 3. Reactor Control - from the standpoint of the maximum pressures attained it is conservative to assume that the reartor is in manual control. If the reactor were in-automatic control, the control rod banks would move prior to trip and-reduce the severity of the transient.
4 Steam Release - no credit is taken for the operation of the sti 'p system or steam generator power-operated relief valves. The sti en-erator pressure rises to the safety valve setpoint where steam re.wse through safety valves limits secondary steam pressure ;t th; ;;tph wa4ue,
- 5. Pressurizer Spray and Power-Operated Relief Valves:- tw n::: 'c- bet"-
e= ;r.d ;.eai,,,um , c a u w i , . i., f emm on . c onei n cu,
- a. Full credit is taken for the effect of pressurizer spray and oower-operated relief valves in reducing or limiting the coclant pressureg 61ssuriser saafety valves are also available, m nh 'm% wJwy system pmauce. c ne,
- b. No credit is taken 'for the effect of pressurizer spray and power operated relief valves in reducing or limiting the coolant pressure,
. Erusurher stafetyvalvesareg;gegg. qh h pr m g h m g Lwhi 6. Feedwater Flow - main feedwater flow to the steam generators is assumed to. -Gro r be lost at the. time of turbine trip. No credit is taken for auxiliary Md feedwater flow since a stabilized plant condition will be reached beforea, ' P S*" auxiliary feedwater initiation is normally assumed to occur. However, the auxiliary feedwater pumps would be expected to start on a trip of the main feedwater pumps. The auxiliary feedwater flow would remove core decay l
heat following plant stabilization, t Reactor trip is actuated by the first Reactor Protection System trip set-7. point reached. Trip signals are expected due to high pressurizer pressure, overtemperature AT, high pressurizer water level, and low-low steac generator water level. No credit is assumed in this analysis for a reactor trio due - to a . turbine trip. Plant characteristics and initial conditions are further discussed in Section The. Mmum sacendocy system pre.asura case. is Except as discussec' above, normal rea : tor coolant system and Engineered Safety Systems are not required to function.1 5: n 21 c2rer 2 e presented in which j pressurizer spray and power" operated relief valves are assumed, but the : l ' Siting : eses ah:re there #"r.cti=: r e ~ + -_ r d : r: :1 = ;r= =te6 l Thiais b de.6y re.cdoc bip en hgk era.uurigg, pra.tsurg,, l The Reactor Protection System may be required to func. tion following a turbine
- l. trip. Pressurizer safety valves and/or steam generator safety. valves may be' L required to open to maintain system pressures below al%wable limits. No i single active failure will prevent operation of any systera required to function.
L A discussion of ATVT considerations is presented in Reference 2. l l i 15.2-6 N M"" -
.The safety valves are assamed to be full open at an accumulation.
- pressars-3% above-the' adjusted lift setpoint which is, in turn, 3%
above!the nominal-lift sotpoint.
+
-e f
i l 4 1. L. l
- l. - . :. -
j l
%m Oce e cbcked pcse.
Tt transient responses for a turbine trip from 100 percent of full power oper tion are shown for four cases: two cases for minimum reactivity feectack and tw cases for maximum reactivity feeocack (Figures 15.2.3-1 through 15.2< -8). The cal lated sequence of events for the accident is shown in Table 15.2. V1.
/
Figures 15. 3-1 and 15.2.3-2 show the transient responses for the total loss of steam loa with a positive moderator temperature coef ficient assuming full credit for the ressurizer spray and pressurizer power-operated re}fef valves. No credit is tak for the steam dump. The reactor is tripped by'the high pressurizer pressu e trip channel. The minimum DNBR remains well above the limit value. The p ssurizer safety valves are actuated for/his case and maintain system oress e below 110 percent of the design value. The steam generator safety valved limit the secondary steam conditic'ns to saturation at the safety valve setpoint. Figures 15.2.3-3 and 15.2.3- show the response fer,t e total loss of s:eam load with a large negative mo rator temperature coefficient. All other plant parameters are the same as the ove. The DNBR increases throughout the tran-sient and never drops below its i itial value./ Pressurizer power-operated relief valves and steam generator fety valvss prevent overpressurization in primary and secondary systems, resp tivel[ The pressurizer safety valves are not actuated for this case. ( In the event that feedwater flow is nyf. erminated at the time of turbine trip for this case, flow would continue j,ider tomatic control with the reactor at a reduced power. The operator wo fd take a tion to terminate the transient and bring the plant to a stabilized ondition. f no accion were taken by the op-erator, the reduced power oper ion would con nue until the contenser hotwell was emptied. A low-low stea generator water i vel reactor trip would be gen-erated along with auxiliar feedwater initiation ignals. Auxiliary feedwater would then be used to re. e decay heat with the r ults less severe than those presented in Section 15 .7, Loss of Normal Feedwate Flow. The turbine trip acc' ent was also studieo assuming the lant to be initially operating at full p er with no credit taken for the pres rizec spray, pres-surizer power oper ted relief valves, or steam dump. The actor is tripped on the high pressur' er prt.ssure signal. Figures 15.2.3-5 and .2.3-6 show the transients wit a positive moderator temperature coefficient, he neutron flux remains essen ally constant at full power until the reactor is ipped. The DNBR never es below the initial value throughout *.he transient. In this case, the pressu nier safety valves are actuated and maintain system pres re below 110 perce t of the design value. Figure 15.2.3-7 and 15.2.3-8 are the transients with maximum reactivity feed-back ith the other assumptions being the same as in the preceding case. Ag 'n, the ONBR increases throughout the transierit and the pressurizer safec v ves are actuated to limit primary pressure. Reference 1 presents additional results of an analysis for a complete loss of heat tink including loss of main feedwater. This analysis shows the overpressure protection that is afforded by the pressurizer and steam generatur safety valves. 15.2-7 1986 Locate
pesults The transient response for a turbine trip from 102 percent of full power is presented for the maximum Main Steam System pressure case and the maximum primary system case. Since the transient response is virtually identical, except for the difference in primary pressure response, for the two cases, the system response is presented for the maximum Main Steam System pressure case, and only *.he primary pressure response of the maximum primary system pressure case is presented.
'he calculated sequence of evants for the accident is shown in Table 15.2.3-1.
Figures 15.2.3-1 through 15.2.3-5 show the transient response follow-l ing a. turbine trip for the maximum Main Steam System pressure case. rull credit is taken for-the pressurizer sprays and pressurizer power-
-operated relief valves. No credit is taken for t7e steam dump. The reactor is tripped by the OTAT trip channel. The steam generator
'_~ safety valves limit the Main Steam System pressure below 110 percent of the design value. The transient response following a turbine trip for the maximum primary system pressure case is virtually identics1 as that pres:cnted for the maximum Main Steam System pressure case in Figures 15.2.3-1 ' through 15.2.3-4, except for the primary system pressure response-provided in Figure 15.2.3-5. The primary system pressure response for the maximum-primary system pressure case is presented in Figure 15.2.3-6. No credit is taken for the pressurizer sprays and pressur-izer power-operated relief valves. No credit is taken for the steam dump. Tt.e reactor is tripped by the high pressurizer pressure trip function. The steam generator safety valves limit the Main Stear ! System pressure below 110 percent of the design value. The L pressurizer safety valves limit the primary system pressure below L10 percent of_the design value.
15.2.3 3 Environmental Consecuences 1 The radiological consequences resulting from atmospheric steam dump will be ! less severe than the steamline braak event analyzed in Section 15.1.5.3 since l no fuel damage is postulated to occur. l 15.2.3.4 Conclusions Results of: the analyses, including those in Reference 1, show that the plant-
- design-is such that a turbine trip presents no hazard to the integrity of the
.RCS or the main steam system. Pressure-relieving devices incorporated in the two systems are adequate to limit the maximum pressures to within the des _ign' limits.
N f the core is maintained by cperation of the D c t ; c h v i.w oh System, i.e. , t.ne DC M1 he maintain-d :.Luve 1.ne limit value. The applicable acceptance criteria as ' h 3 C ivo IL4LL.htve been met. The above
~ -i K65 the ability of tbt N555 to safety wi s t:M g 15.2.4' INADVERTENT CLOSURE'0F MAIN STEAM ISOLATION VALVES d Inadvertent closure of the main steam isolation valves would result in a MM POojt.
l turbine trip. Turbine-trips are discussed in Section 15.2.3. i 15.2.5 LOSS OF CONDENSER VACUUM AND OTHER EVENTS CAUSING A TURBINE TRIP Loss of condenser vacuum is one of the events that can cause a turbine trip. Turbine trip initiating events are aescribed in Section 10.2.2. A loss of condenser vacuum would preclude the use of steam dump to the condenser. However, since steam _dumo is assumed not to be available in the turbine trip analysis, no additional. adverse effects would result if-the turbine trip were caused by_ loss of condenser vacuum. Therefore,. the analysis- results and conclusions contained in Section 15.2.3 apply to loss of condenser vacuum. In addition,- analyses for the other possible causes of a turoine trip, as-listed -in Section
.10.2.2 are covered by Section.15.2.3.- Possible overfrequency effects'due to a turbine overspeed condition are discussed in.Section 15.2.2.1 and are not a-concern for.this type of event. .
15.2.6- LOSS OF NON-EMERGENCY AC-POWER TO THE STATION AUXILIARIES 15.2.6.1- Identification of Causes and Accident Descriotion A. complete Icss of non-emergency AC~_ power may result in the loss of a' puer to the plant auxiliaries, i.e. , the reactor coolant pumps, condensate :.e'.ps . etc. The loss of-power may be caused by a complete loss of the offsito grid accompanied by a turbir.e generator-trip-at the station, or by a loss of the onsite AC distribution system. f 15.2-8 1984 Lpdate f
-ie- e - .. %,-_ , . . - , - - , . , _ , -
,.~u ,,--e-. - g +-rw,,w -
_ - . . .. . . - . - . ~ . . ~ - - - - _ - ~ . . - . - - - - . ~ . . . - -.-.~...n.~- __--- - . Inadvertent' closure of the main steam isolation valves (MSIV) would result-in a~ transient response similar to that of a turbine trip. The closure of the MSIVs would isolate a smaller volume of steam piping. than-a turbine trip, which.would tend to cause a transient more severe than a turbine trip event. However, the longer closing time of the MSIVs, relative to the turbine stop valvo closure time, offsets the ef fects of the smaller st.eam piping volume, and therefore the MSIV closure event is less severe than a turbine t rip event. Turbine trips are discussed-in Section 15,2,3, l l I i 6 2
, ,- - ,, ..n -
~--w. n n ~ . , - , - , . , , , , - . ,
i 15.2.8 FEEDWATER SYSTEM PIPE BREAK 15.2.8.1 Identification of Causes and Accident Description A major feedwater line rupture is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell side fluid inventory in the steam generators. If the break is postulated in a feedline between the check valve and the steam generator, fluid from the steam generator may also be discharged through the break. Further, a break in this location could preclude the subsequent addition of auxiliary feedwater to the affecteu steam generator. (A break upstream of the feedline check valve would affect the Nuclear Steam Supply System only as a loss of feedwater. This case is covered by the evaluation in Section 15.2.7). Depending upon the size of the break and the plant operating conditions at the time of the break, the break could cause either a RCS cooldown (by excessive energy discharge througc the break) or a RCS heatup. Potential RCS cooldown resulting from a secondary pipe rupture is evaluated in Section 15.1.5. Therefore, only the RCS heatup effects are evaluated for a feedwater line rupture. A feedwater line rupture ceduces the ability to remove heat generated by the core f rom the RCS for the .'ollowing reasons:
- 1. Feedwater flow to the steam generators is reduced. Since feedwater is i subcooled, its Icss may cause reactor coolant temperatures to increase prior to reactor trip.
- 2. Fluid in the steam generator may be discharged through the break, and would then not be available for decay helt removal after trip.
- 3. The break may be large enough to prevent the addition of any main feed-water after trip.
An Auxiliary Feedwater System is provided to ensure that adequate feedwater will be available such that: i
- 1. No substantial overpressurization of the RCS shall occur.
- 2. Sufficient liquid in the RCS shall be maintained in order to provide adequate decay heat removal.
Section 10.4.7 contains a description of the Auxiliary Feedwater System l interfaces. The severity of the feedwater line rupture transient depends on a number of system parameters including break size, initial reactor power, and credit taken for the functioning of various control and safety systems. A number of cases t of feedwater line break have been analyzed. Based on these analyses, it has l been shown that the most limiting feedwater line ruptures are the double ended rupture of the largest feedwater lineg accm ,i % um f u l l pe.ua ai m anu w i muut 4cas -ef of f;itc pc.ag with no credit taken for pressurizer control. Thlie-casey are analyzed below. IS l t 15.2-15 3/90
The following provides the necessary protection for a main feeawater rupture:
- 1. A reactor trip on any of the following conditions:
( a. High pressurizer pressure. ( b. Overtemperature aT.
- c. Low-low narrow range levei in any steam generator. )
i
- d. Safety injection signals from any of the following: l l
- 1. 2/3 low steam line pressure in any loop.
- 2. 2/3 high containment pressure p
(Chaoter 7 contains a description of the actuation system).
- 2. An Auxiliary Feeawater System to provide an assured source of feeowater to the steam generators for decay neat removal. (Section 10.4.7 contains a description of the Auxiliary Feedwater System).
A major .feedwater line rupture is classified as an ANS Condition IV event, a l limiting fault. See Section 15.0.1 for a discussion of Condition IV events. 15.2.8.2 Analysis of Effects and Consecuences Method of Analvsis l RETRAN o2 6 i A detaileo analysis using the 14F+**T Code (Reference J) is performed in orcer to determine the plant transient following a feuJwater line rupture. The coce L describes the plant thermal kinetics, RCS including natural circulation, pres-surizer, steam generators and feeawater system, and ccmoutes pertinent sariables including the_ pressurizer pressure, pressurizer water level, anc reactor coolant average temoerature. !' The cases ana'yzed assume a douole ended rupture.of the largest feedwater cipe at full power. Major assumotions made in the analyses are as foliows: f
- 1. The plant is initially operating at 102 percent of the NSSS power rating.
'f.o
- 2. Initial reactor coolant average temperature is 3<d*F above the nominal.
value and the initial pressurizer pressure is 30 osi above its nominal l value. I
- 3. No. credit is taken for the pressurizer power-operated relief valves or pressurizer spray.
4 Initial cressurizer level is at the nominal programmad value plus 9 per-cent 4,n:rre,7; a r..,. , 4n initial .w. steam generator re,,itec ,+- - water level n n.catn, is 2,at the ema .s. nominal
- : :i value p ua 7 p_, ,
minus 8.X percent. in tr.: int:: teer ;e t-ste-s. O 15.2-16 3/90 l l' l
- 5. No credit is taken for the high pressurizer pressure reactor trip.
- 6. Main feedwater flow to all steam gentrato,s is assumed to be lost at the time the break occurs (all main feedwater spills out through the break).
- 7. The worst possible break area is assumed. This maximizes the blowdown discharge rate following i.he time of trip, which maximizes the resultant heatup of the reactor coolant, fqktg --
'm 4 8. A conservative feedline break discharge quality is assumed prior to the M time the reactor trip occurs, thereby maximizing the time until the trip setpoint is reached. After the trip occurs, a saturated liquid discharge 1 is assumed until all the water inventory is discharged from the af fected steam ganerator. This minimizes the heut removal capability of the J
t affected steam aenerator. hgk to d.om%-b prt.13ure
- 9. Reactor trip is assumed to tee initiuted on t. 'ec h h vel trip setpvint in t he A 4 team generstet -10 pg7cggt gf n.,ppew M".ge Op3 g
. 54eby We. m
- 10. The Auxiliary Feedwater System is actuated by the low-low steam generator ) .
4 water level signal. The Auxiliary Feedwater System is assumed to l initially supply a total of 335 gallons per minute (gpm) to the three
- 2. unaffected steam generators, including an allowance for possible spillage through the main feedwater line break. A volume of 38 ft3 is assumed for I each feedwater line which must be purged before the relatively cold ,
(110 F) auxiliary feedwa+er enters the unaffected steam generators.
- Fifteen minutes following the low-low level signal, operator action was ,
assumed to terminate auxiliary feedwater flow from spilling out the break. , This increases the auxiliary feedwater supply to 473 gpm to the three unaffected steam generators.
- 11. No credit is taken for thermal energy deposited in RCS metal during the RCS heatup.
12 No credit is taken for charging or letdown. M . L13. Steam generator heat transfer area is assumed to decrease as the shell [ t side liquid inventory decreases. I 13 J#, Core residual heat generation is assumed based upon the 1979 version of ANS-5.1 (Reference 5). ANSI /ANS-5.1-1979 is a conservative representation of the decay energy release rates, ti JF. No credit is taken for the following potential protection logic signals to mitigate the consequences of the accident:
- a. High pressurizer pressure,
- b. Overtemperature AT.
- c. High pressurizer level.
i H gh Cea+ d -ent e-e:c=. 15.2-17 3/M-
.7 g..- , . . . . . - - - - - , .-.- .. _- _.. - . - _ , ... . -
~l
-Attachment 1:
~
8' , The flow exiting the steam generator through the break is , determined using the Moody (subcooled) and Extended-Henry (saturated) correlations for choked flow. l Attachment 2:
- 10. .The-Auxiliary Feedwater System is actuated by the high containment 1 pressure safety injection-signal. Tne amount of. 1
' auxiliary feedwater delivered to the faulted and intact: steam generators is a function of the individual generator pressures, l A volume of 40 ' f 2 t -is assumed for each feedwater line which must be purged of relatively hot main feedwater before-the cold (138*F) auxiliary feedwater enters the steam generators. Two minutes into the transient,. operator action is assumed to-. ,
terminate-auxiliary.feedwater flow to the faulted generator.
~
l This increases.the auxiliary feedwater supply to the three-unaffected steam generators, but reduces the total flow delivered to.all four' generators. Therefore, this eirly operator action, ,
- in.effect, degrades the secondary side heat removal capacity.
)
4 4 J m
+ w w - =rM- =_-- y .,,,w..e e-%-as- m - r, ,m-- y- m.- **e 4 'M-w' g
ig 'g. 10% of the tubes in eacn steam generator were assumea to ce plugged. This was cone to support uniform olugging of the steam generater tubes uo to a , m 9tg\,xa. &aximum DMc%et,d level\ of 10% for any steam generator. F Receipt of a low-low steam generator narrow range level signal in at least one steam generator starts tne motor-criven auxiliary feeawater pumos, *nicn tnen f deliver auxiliary feedwater flow to the steam generators. The turoine driven auxiliary feeawater pump is initiated if the low-low steam generator water level signal is reached in at least two steam generators. Ste:ilarly, receipt of a low staa. line pressure signal in at least one steem line initiates a steam line isolation signal wnich closes the main steam line isolation valves in all steam lines. This signal also gives a safety The injection signal wnich amount of safety injection initiates flow of boratea water into the RCS. -- { flow is a function of RCS oressure. Emergency operating procedures following a secondary system line rupture call for the following actions to be taken by the reactor operator:
- 1. Isolate feeawater flow spilling out the break of rupturea steam generator and align system so level in intact steam generators recovers.
- 2. Stop the safety injection pumps if:
- a. Wide range reactor coolant pressure is increasing,
- b. Pressurizer water level is on span,
- c. The RCS is adecuately suoccoled, and
- d. Steam generator narrow rance level indication exists in at least one steam generator or suf ficient auxiliary feecwater is being injected into the steam generators to provide an aceouate neat sink.
Lt.% beheed 2. Subsequent to r ecovery of level in the intact steam generators, the olant operating proceaures will be followed in cooling the olant to not snutcown conditions. Plant characteristics and initial conditions are further discussed in Section 15.0.3. No reactor control systems are assumed to function. The Reactor Protection System is required to function following a feeowater line rupture as analyzea A nere. No single active f ailure will prevent operation of this system. aiscussion of ATWT consicerations is presented in Reference 2. 1 The engineered safety systems assumec to function are the Auxiliary Feeawater System and the Safety Injection System. For the Auxiliaryit:= Feedwater System,
;=x A n:n the worst case configuration nas been used, i . e . , it-- n cr i-f e :r !;rj '::r~ T ' i ng th: ; rude the turoine-driven auxiliary feeawater pumo nas been assumea to f ail.{ th; -o ~nr 9 cr m: i:;;??-
n;c; !:;_ m3 ; ;, i ' i m vn
*10 :r M5 ;- t: th: . r: == :t:r g;... ..
.u_ c:!1._,,,3 ,.: r:te : n k n w . .v i ne o m i,;rj n:.__ a iivw
- " in: * *
, :n ;- ..d, m. i .~ ~ ;. , . ;r i . c ,, ;;; :;! . ; r ' M ga t: " " -e n;n 3e rm . ;n c ; . Only one train of safety injection nas been assumed to De 15.2-18 3/90
- . , ,. --_._.:...- - . . - . - - . . . - _ - - . . . . - .. - - . - . . - . ~ . . - . - - . . - . . . . ~
l Attachment'1:
- 16. Safety injection actuation occurs upon the receipt of a high containment pressure signal. The amount of safety injection flow ,
delivered to the Reactor Coolant System is a function of RCS pressure. A high-high containment pressure signal generates a main steam line isolation signal which closes the steam litie ' isolation valves in all four steam lines.
-Attachment 2:
- 3. _Stop the reactor coolant pumps when containment pressure reaches the high-high pressure setpoint..
r I t L _ ~ +- . , . . .. , . . , , .._ __ _ . . - . . . . , _ _ . - . _ . - - , -
svailable. A detailed descriotion ano analysis of the Safety Injection 3 stem / is oroviceo in Section 6.3. The Auxiliary Feedwater System is cescrioeo in Section 10.4.7.
%%6s
~
er sea.rie c4 Se. w.acwr cook + Pumes e cue -!tne;. of4 it ;: er tnere will be a flow coastcown until n ow in the loops reaches tr.e natural circulation value. The natural circulation cso-ability of the RCS has been shown in ".ection 15.2.6, for the loss of AC non-emergency cower transient, to be suf ficient to remove core decay heat f ollowing reactor trip. Pump coastdown characteristics are demonstrated in Section* 15.2.1 r: 15.3.2 for : gic c-c multiple reactor coolant pump trips, e- r::ti sif. L u d.,A W 'b
- d I.
Results Calculated plant parameters following a major feedwater line rupture are similar to that given in Figures 15.2.3-1 through 15.2.8-3{.9 Result: ': - a care ith ef f:!t: peu:r reat!aol :r: pr:::nt: c!gur:; 1;. 2. 0 . . m -gn 15.2.0 5. 's,wii5 Tm Los .536 -nere affsiti gn=ri ia Ivas aE pi c aca u s ' p uur.< u 'oc +- : ;n.15,2.3-11. *he calculated sequences of events 4-<
"a*4 r : m:!;::c i* :f ar ': **st given in Table 15.2.3-1. '"e etu - -
- nct- r: '; : c!';nt!y : : :ansec.ati a transi:nt with a 7 ysm mecc. :t;-
...m....m.. ___2,= -.
Qt SA%eCA The system response I~= following the taeewater line rupture is similar for oath cases analyzed. Results presented in Figures 15.2.8-3 and 15.2.8-5 (with off-site power available) ana Figures 15.2.8-8 and 15.2.8-10 (without of fsite power) show that pressures in the RCS and main steam system remain below 110 percent of the respeci dve design pressures. Pressurizer presstre increases ! until re-actor trip occurs on low-low steam generator narruw range level. Pressure -then decreases, due to the loss of heat input, until the Safety Injection system is actuateo on icw steam line pressure in the ruptureo ' coa. Coolant exoansion occurs due to reduceo heat transfer capaoility in the steam generators. The pressurizer safety valves open tc maintain primary pressure it an acceptacle value. Addition of the safety injection flow aids in cooling down the artmary ano helps to ensure tnat suf ficient fluio exists to reeo tne core covereo with water. IIjw. s 1 15.2.0-1 2-d 15.2.2 # ^t th:t fu!!vwin3 i nau war trip bue pisai~~
- s ... . ;; rueca+4-=1 RCS pressure will be maintained at the safety valve setpoint until safety in-jection flow is terminated by the coerator, as mentioned in Section 15.2.3.2.
The reactor core remains covereo with aater throughout the transient, as -ater relief due to thermal expansion is limited by the neat removal capability of l the Auxiliary Feedwater System and makeup is provided by the Safety Injection System. l The major cifference between the two cases analyzed can be seen in the clots of hot 'ano cQId leg temperatures, Figure 15.2.8-4 (with of fsite power available) and Figure 15.2.8-9 (without of f site power). It is apparent f rom the initial portion of tne transient (s400 seconos), that the cast without offsite power ' results in nigner temperatures in the not leg. For longer times, however, the case with offsite pcwer results in a more severe rise in temperature until the coolant pumos are turned of f' and the Auxiliary Feedwater System is realigneo. 1 15.2-19 -W M-
. - . - ~ _ - - - . - - . - - , _~ ~. ~ -. - -.- ---
i
'l i
- Attachment 1:
-The:DNBR calculation for.this-accident is performed with the VIPRE-01 computer code (Reference 7) using the Statistical Core Design procedure described in Reference 8. DNBR is a concern for this transient _because the assumed Icss of offsite power causes a-reactor coolant pump coastdown. Because of the loss of secondary heat sink, the RCS temperature is somewhat higher than normal when the coastdown occurs. Since the~ loss of offsite power is assumed to occur coincident with reactor and turbine trip, the amount of.heatup prior to the coastdown would be limited by the overtemperature AT trip function. This trip-setpoint is reduced by RCS heatup. The heatup allowed by the overtemperature AT trip function 1s-conservatively
- assumed in the DNBR calculation. _Although either high containment pressure safety injection or steam generator low-low narrow range level is expected to trip the reactor for this transient, no credit is j taken for'these trip functions in the DNBR calculation. The Technical !
Specification limits on axial flux difference ensure that the DNBR remains above the limit value for the feedwater line break transient, including the effects of loss of offsite pcwer and reactor coolant pump coastdown Attachment 2: Pressurizer pressure, shown in Figure-15.2.8-3, increases initially on the loss-of heat sink, then decreases due to the blowdown of the faulted steam generator and the loss of heat input following reactor trip. As the faulted generator blows dry, the overcooling phase of the transient is terminated and pressure again begins to increase. The auxiliary feedwater system heat removal capacity turns this pressure increase around before the pressurizer safety valves are,even challenged. The pressurizer safety valves relief capacity is adequate
.to mainta ' primary system pressure below 110 percent ofedesign
' pressure.
-Figures-15.2.8-6 shows'that the Main Steam System-pressure in the intact generators holding fairly steady near the safety valve lift n
'.setpoint, well below 110 percent of the design ~ pressure.
i ! Figure 15.2.8-9 shows that the calculated'DNBR remains above the limit-value,'thus ensuring that the integrity of the core is maintained in the short-term. The adequacy of the auxiliary feedwater system'to L remove decay heat,- as shown by Figure 15.2.8-2, ensures the integrity of the core in the long tern.. L l-
Daw 7
'The-pressurizer fills for both the with and without power cases due to the increased coolant expansion resulting from the oecay heat and stored energy. ]
Hence, water is relieved for both the with and without power cases. As i toreviously stated, however, the core remains coverea with water for both cases.a ( 15.2.8.3 Environmental Crnsecuences The feedwater line break with the most significant consequences would be one that occurred inside the containment between a steam generator and the feed-line check valve. In this case, the contents of the steam generator wow d be released to the containment. Since no fuel failures are postulated, the
- radioactivity released would be-less than that for the steamline break, as analyzed in Section 15.1.5. Furthermore, automatic isolation of the contain-
' ment would further reduce any radiological consequences from this postulateo ! accident. 15.2.8.4 Conclusions Results of the analyses show that for the costulated feedwater line rupture, the assumed Auxiliary Feedwater System capacity is adequate tu remove decay heat, to prevent overpressurizir, the RCS, and to prevent uncovering the re-I actor core. Radioactivity doses from the postulated feedwater line rupture are ' less than those previously presented for the postulated steam line break. All applicable acceptance criteria are therefore met. . l l 15.2-20 -#90- _ _ _ _ _ _ - _ _ _ _ . - . _ _ _ _ _ _ _ -- . . - .-. . - . - -= , -
REFERENCES FOR SECTION 15.2
- 1. Mangan, M. A. , " Overpressure Protection for Westinghouse Pressurized Water Reactors", WCAP-7769, October 1971.
- 2. " Westinghouse Anticipated Transients Withour Trip Analysis",
WCAP-8330, Auoust 1974.
- 3. Burnett, T. W. T. , et al. , "LOFTRAN Code Description", WCAP-7907-P- A (Proprietary), WCAP-7907- A (Non-Properietary).
- 4. Chelemer, H. , Boman, L. H. , Sharp, D. R. , " Improved Thermal Design Procedures", WCAP-8567, July 1975.
- 5. ANSI /ANS-5-1-1979, "American National Standart! for Oecay Heat in Light Water Reactor.", August, 1979.
$nSeFD hom obched page. 9 15.2-21 1936 Update
. . . _ _. - - - __ _ . . _ _ . . . _ _ . . _ .._.- . . _. . = . _ __. ..-.._.___.-._ _ _ m _ ~_. . ._._
h RETRAN-02: A Program for-Transient Thermal-Hyd'raulic
- 6. .EPRI, "
Analysis of Complex Fluid Flow Systems", ffRT NP-1850-CCM, Revision 4,-November 1988.
- 7. EPRI, "VIPRE-01: A Thermal-Hydraulic Code:for Reactor Cores",
EPRT NP-2511-CCM-A, Revision 2, July 1985.
- 8. " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01", DP0-NE-2004P, Revision 1, February 1990.
h
=
N
+
5
?
_ _ - _ _ . . _ _ _ _ __ .m. _ .,. _ , , . ~ . , . . _ , , , , _ _
Tabl e '15. 2. 2-1 Parameters for loss of External Load Dose Analysis
- 1. Data and assemptions used to estimate radioactive sources from postulated accidents. *
- a. Power level (MWt) 3565 1
- b. Percent of fuel defective 1 I
- c. Steam generator tube leak rate prior to and I during accident (gpm) l 1
- d. Offsite power Not available l 1
- e. Reactor coolant activ.ity Table 15.0..tf-2 I 9
- 2. Data and assumptions used to estimate activity l released. '
ol
- a. Iodine pa.;ition factor for steam releases 0./
- b. Release from steam generators (lbm)
(0-2 hrs) 508,000 (2-8 hrs) t,200,000
- 3. Dispersion data
- a. Distance to exclusion area boundary (m) 762
- b. Distance to low population zone (ft) 8839
- c. X/Q at exclusion area boundary (sec/m3 )
(0-2 hrs) 9.0E-4
- d. X/Q at low population zone (sec/m3 )
(0-8 hrs) 8.0E-5
- 4. Dose data
- a. Method of dose calculations Appendix 15A
- b. Dose conversion assumptions Appendix 15A
- c. Doses (Rem) see. Tale,15.0.7-l Q } Exclusion area boundary ~
Whole body - Thyroid 1.9E-01 Low population zone Whole body 8.5E-04 Thyroid 8.1E-02
-V % ':p i=
Table 15.2.3-1 (Page 1 of 5) Time Secuence Of Etents For Incidents Which Cause A Decrease {g In Heat Removal By The Secondary Systems b ?.krgv Accident Event Time ( ec) Turbine rip
- 1. With p .surizer Turbine trip, loss 0.0 control inimum of main feed flow reactivity feedback) i Initiation of steam 7.0 release from steam gen j erator safety valves /
High pressurizer p ssure 7. 3 reactor trip set int reached Rods begin t drop 9.3 Pear, pres rizer 11.0 eseure ccurs Mi DNBR occurs 11.5
- 2. With pressurizer T bine ip, loss 0.0 control (maximum main f d flow reactivity feedoack)
Peak pressurt er 7.0 pressure occur Initiation of ste i 7.0 release from steam . n-erator safety valves Low-low steam gen- 52.6 erator level reactor trip setooint reacned Rods begin to drop 56.6 Minimum DNBR occurs (1) (1) R does not cecrease below its initial value. 1986 Update
l I.
-Accident Event Time (see)
Turbine. Trip
. l. Maximum Secondary Turbine Trip, loss of 0.0 System Pressure Case main feed flow Pressurizer PORVs lift 3.4 Steam Safety Valves lift 5.4 l Overtemperature delta T 13.0 setpoint reached Control rod insertion 14.5 begins Peak secondary system 18.7 pressure occurs 1
- 2. -Maximum 1 Primary Turbine Trip, loss of 0.0 l
_ System Pressure Case. main feed flow High pressurizer pressure 4.5 L~ setpoint reached-Control rod insertion 6.5 begins Pressurizer Safety Valves 6.8
- 11.f t :
. Steam Safety Valves lift 7.4 Peak; primary system 7.7
-pressure occurs I
i lL v , e s ,. - ,-n , , -,---,r -
~ .
bed % Table 15.2.3-1 (Page 2 of 5) j me Seouence Of Events For Incidents Which Cause A D. rease In Heat Removal 8v The Secondary System
)
Accident Event Time (Sec) Turbine Trip
- 3. Without pressurizer Turbine trip, I ss 0. 0 l control (minimum of main feed ow j reactivity feeoback)
High press izer pressure 3. 7 i reactor t p setpoint reached j l ds egin to drop 5.7 I iation of steam 6.5 . le ce from steam gen-erato safety valves Peak pr ssurizer 7.0 pressure occurs Minimum DN R occurs (1)
- 4. Without pressurize Turbine trip, loss 0.0 control (maximum of main feed t ow reactivity feeo ck)
Hign pressurize pressure 3. 7 reactor trip set int reached Rods begin to urop 5.7 Initiation of steam 6.5 release from steam gen erator safety valves Peak pressurizer 6. 5 pressure occurs Minimum DNBR occurs (1)
- 11) ONBR oces not decrease below its initial value.
1986 Update
!= Table 15.2.3-1 (Page 4 of 5) Time Seouence Of E,ents For Incidents Wh'ch Cause A Decrease In Heat 2emoval Bv The Secondarv System Accidant Event Time (Sec) Core decay heat plus pumo - 1700 2 Loss of Normal Feeawater Flow (cont'd) heat cecreases to auxiliary i feeowater heat removal ( capacity ( Feeowater System vipe creak I 4 With offsite power Main f esidline ruptura uccurs 10 Ru%cc 1. ( w* available M Low-low steam generator level 25 ( NOE reactor trip setpoint reatnec in ruptuted steam generator f' Rods begin to drop 39
/
Auxillary feecwater is started 95 Feecwater lines are purged and 212 d cold auxiliary feecwater is g deliverec*;c two of the three 3 intact steam generators ( Low steam line pressure 248 l ( setcoint reacned in rupturec ( steam generator (' I Ail main steam line isolation 255 valves close C Feeowater line is purgeo and 474 < cold auxiliary feeawater is 's delivered to the third I intact steam generator G 4 Pressurizer water relief cegins 609 < Steam generator safety valve 711 ( setpoint reatned in two of the 4 three intact steam generators
}
Steam generator safety valve 1145 ( reacned in third intact steam generator _. 02/00_
. -._ . _ _ . _. < . . _ . . . _ _ . . .. _...___._m._=_...._.. --.- _ __ _.- ____ ___.___.
'Feedwater line break to SG B 0 Safety injection on- 10.06 high containment pressure Reactor trip on high .
IJ.06 containment pressure SI Turbine trip on reactor trip 10,21 Reactor coolant pumps tripped 15 Steam line isolation on hi-hi 15 containment pressure , Safety injection terminated 70 . Motor-driven CA pump delivers 76 flow CA to faulted generator isolated 120 [ SG B boiled dry 133-l. l Core decay heat decreases to ~1000 auxiliary feedwater heat. ' removal capacity
.?
End of simulation 2000
-;v.
l i l I%
' h, w
i
--'t:a' -r-- -
DAk P, Table 15.2.3-1 (Page 5 of 5) Time Seouence Of Events For Incicents which Cause A Decrease
-In Heat Removal By The Secondary System Event Time (Sec)
Accicent
\
Feecwater Systte Pipe Break Core decay heat plus pump heat - 50
- 1. With offsit power decreases to auxiliary feeowater available(con (d) heat removal capacity Main feedline rupture occ s 10
- 2. Without offsite powet s i
available N level 35 l g low-low steam generat reactor trip setpoi reacned
- .g
-. in ruptured steam enerator Rods begin to rop 39 A liary eedwater is started- 95 Feec .er lines are purgeo and 212 col x;11ary feedwater is i
de ve t, two of the three l # tact s am generators Low steam 1 ne pressure 285 f ' setpoint rea ec in ruoturec steam generato All main steam i e isolation 292 l valves close l l 474 Feecwater line is pu ged and cold auxiliary feedwater is delivered to the thirc intact steam generator Pressuruer water relief b gins 751 Steam generator safety valve 785 l setpoint reached in two of the l three intact steam generators l Steam generator safety valve 211 l reatneo in third intact steam generator Core decay heat decreases - 1275 to auxiliary feedwater l heat removal caoacity l i / 03/90 \ l
1.2000 e E 5 w
- . 0000 -ll.)
*O 30000 - - 4 m
- E .50000 - -
a 3 . c .40000 - I e w - i G 20000 -
+
E i
.0 o e o o -
8 8 8 8 5
\ o 6
.o ~ 6
. o. 8 T!HC (5 s 2600.0 '
, 2500.0 - l- '
a S 2400.0 - -
+
} '<
i E
~
2300.0 - -
= :<:; 2200. 0 - -
v 1
~u E~ 2100.0 - - l S
=
2000,0 - f
!$00. 0 - - .
1800.0 8 8
- g 8
- a. a o
o o o. .o . - TlHE ($[C] 1600.0 T 2 1500.0 - -
+
i ! 3 1400.0 - 1 > >
,_ 1300.0 - -
~
i 1200.0 " 7 T 1 0 -
$ ':t 900. --
p L $ 800 00 - - a
? 7" .C0 y
l 0.00 . 8 8 8- 8 8
- o. o o 6 o
- o ~ o. so o. -
ilHE tSEC T' INE TE!? I'.'ENT Ent
? .ESSUR :IR SPPAY AND P0t.*ER PERATC R.". IT VALVES, s
..IN!.'t.M RIACT".'IT;" FIDBACK
. mimer McGUIRE NUCLEAR STATION Figure 13.2.3-1
'936 UPDATE
R 400.00 : i [ 55J.00 -- .+ 2
$- 180.00 -- ,,
-t :
*G 570.00 -- 4, O9 -
Z 160.00 -
.a 8 '50.00 -- -,
l 540.00 a a 8 8 a . !
*. 5 a, .o, 6 8
- )
e ~ l T!HC (5(C) 640.0 i w 3 3 620.00 - - ... r x
= - 600.00 - - ..
d S*. E9 580.00 - - .. ' 5< g 560.00 - - a w
$40.00 a 8
- a a a d o 8
. o. . - d. -. -
T!HC (5(C) 4.0000 \ 3.5000 - - 4. 3.0000 ~- - . 2 io 2.5000 - - 2.0000 - ..
- 1. 500 l- - .
1.0 0 o
'8* 8 8 E 8
*
- 6 e
s e e s = e TIME 15EC) l " T'RBINE TRIP EE '4ITF. PRESSURIZER SP. Y .03 ?C'4ER i CPERATED RELIET .'LVE5. MINIMLH REACTIV!U FEE 03ACK pai wo McGUIRE NUCLEAR STATION Y e igura 15..>. ;. i 1986 UPDATE l' - . .. . - - . _ - . - . . . - -.
; kt? c.e_ 1.2000 z- .
.i
$. 1.C000 :
t
- 5. g: '60000 - ,
_2-$-.60000
')* B 40000 " '
,20000 " "
* /
g- g a a 8
. s-
~ s, .s.
- s. s TIME (SEC) 2600.
, 2500.0 < "-
-$ 2400.0 " ,
2300.0 "[ \ ""
" $. 2200.0 k+ 2100.0 " '"~
- 2000.0 " "-
a
!$00.0 " "
1800.0 I a .a 8- -
- s -
sw o 8-
._ ~ - =
T!N (SEC)- 1500.0 , \. =- T 2 1400.0 - " ,: 5- -1300.0.- " l~ z - 1200.0" " N. b 1100.0 - '
-- 2 a h! 1000.00 w
O " "
- 3.
g
! $00.00.-
800.00 " ' l j 700.00 " J , L ,4 600.00 g g g-
- 6 o o
. - - Y
-"R3INE TRIP ~ EVENT. L'ITH TINC (SEC)
?RESSURIZER SPRAY AND FC JER-CPERATED RELIET VALVES.
XAXIMUM REACTIVITY FEF.DBkCK A McGUIRE NUCLEAR STATION ms me - M s Ti:;ure 13.0.3 1986-UPDATE
Lt.fLaee_ 600.00 ' y !!0.00 - _ 580'.00 - - -
*u .00 - - -
af 5 w - 1 560. - -- 5 550.00 ' + w
'40 00
- - i i s ?
TlHC (5CC) 640.00 w 3 2 a 620.c0 - - - w I g-w u 600.00 f M i - 580.00 " - V w 560.00 -- " 8 w
'40.00
~
a- a a 8 a
*. d &.
& a.
8-o u . T1HC (5 1 4.0000 / 1.5000 - " 2.0000 " "
=
2.5000 - - 8 2.0000 " T I-1.5000 - I 00 g 8 8 8 8
- o. o o o o d d, d 6
=
8 a ,w a T!HE (SEC1 %*RS NE TRIP E*/ENT WITd PRES 'RI:'ER S? RAY AND ?CWER OPERA ED REI.!ET '.*AL'.*E3,
." .AX LM. , REACT!'/ITY FEE 3ACK
;uu m Tv McGUIRE NUCLEAR STATION W nau. t 3 . .. . ., ..
.e -
Rql % t.2000 , z O
- 1. 0000 <
f 5O- <
.30000 - -
T, E " i 55a.60000 "
^
o 40000 -
= .
w
- =
20000 - 2 <
'O g 8 8 / 8 o
8 o o o 6
- d o 6 o o ~ a. a = -
T!HC (SCC) 2600. , 2500.0 -
?, 2400.0 - - -
+
J. F 2300.0 -
*2 2200.0 - -
CC jo '" 2!00.0 - -
" 2000.0 - "
a
!!00.0 - '
1800.0 e o e o o o 8 o o o o o o
. o d, d d
=
b o ~ . - I C (SCCJ 1500.0 Y, 1400.0 - " 5 1300.0 - - - a - 1200.0 - - WO s-- !!00.0 l m M 1000.00 - - +
= g ---
l w~ 300.00 - " j @ 800.00 - - y 700.00 - - 600.00 f e o o r o e o o o o o 8 o d a, o d a ~ . = - T1HC (SCC 1 l l
- T~GINE .4!? EVENT '41THOC
?RESSURI ER SPRAY OR POWER OFEFATC R.". III VAI"ES .
g M!NIMUM REACT!"! E FEEDBACK aat a si 1.1cGUIRE NUCLEAR STATION Figure 15.2.3-5 1986 UPDATE
pbg 300,00 , 30.00 -- , 3 5,. 580.00 - - - 2" l
* ' $70.00 - + ~
2 560.00 < we 8 w 5 . 00 - +
$40.
8 o 9 5 8 o 2 o 8 N N N 8 b TINC (5'.C)p 640.,00 w ', B 7= 620.CO " -- E z
=- 600.00 - - --
w u w U a9 580.00 w w $50.00 " - 8 w
$ 4 0. 00 8 8 8 8
- o
- o. o o e o o o -
a w a *" T!NC ($CC) 5.0000
+
l 4.5000 - - 4.C000 " -
-3.5000 - -
ee io 3.0000 - -
- 2. CC0
- 2.00- - ?
l 1.' 00 -
' l 2
8
?
8 8 8
* \8e
- a. e o o e o a ~ - a m T!HC (SEC)
TURBINE TRIP EVENT I~'dCUT PRESSURI:'ER SPRAY .GD POWER OPE?ATED REI,IEF VA .VES . MINI.'C.. MODEPa!OR TEE 03 ACK au mvei McGUIRE NUCt. EAR STATION Y : igure !...,.,-e f om J
Dkkg 1.2006 ;
- I 3 1.0000 q +
w 50 0000 - - - E 55 3 =
.600 -- -
o lo 40000 - w
.20000 -
2 3 8 8 8 8 8
- e. o
.' 8 e o, o ~ .o -
ilHC ($[C1 2600.0 m 2500.0 - - -- a S 2400.0 - - - e W 2300,0 - - -.
~
$<a.
~
2200,0 - - -- j '" 2100.0 - - -- o
- 2000.0 - -
a
!S00.0 - - -
1800.0 8 8 8 8 8 o a
- e o,- e o o ~ a o.
a - T!NE t!C 1400.0 w
@ 1300.0 - - -
a> 1200.C - --
$ i t00. 0 - - --
3 ~ 1000.00 - -- cc M
+
43 900.00 - - Ee S 900.00 - - -
+
Ec 700.00 " + I 1 600.00 8 8 8 8 8 e o o o o o N o. e =
/
l IINE (!EC) ) 73.BISE ~'RI? EVDi! IT.40C l PRESSURIZER SPP.AY AND PC'41 l r OPERATED RELIEF VALVES.
'd.AXDG. P . REACT!?!~Y TIEDSACK
- ni mini McOUIRE NUCLEAR STATION l
.%F _igur e L, . ., . 3 ,
l 1986 UPDATE ,
gg 600.00 ' 590.00 -- 4 E
=
g _ 580.00 - r a4 510.00 --
*/
"., Y 2 560.00 H i
. 550.00 - -
w 540.00 e o o o e 8 8 3 8 6
- i s s i e TlHE (SEC' G40.00T w
3 7a 620.00 --
- E
+
9 - 600.00 -- 0S
&9 580.00 - -
5< " w 560.00 - - E w 540.00 g g g g o o o a
*. d d, a d o o u w = -
ME (SEC) 5.0000 4.5000 - - J-4.0000 - - 3.5000 M io 3.000 -- "
+
2.50 0 - -
\
- 2. 00 - -
1 5000 -- I j g g g g 8 6 i s s $ TIME (5Ef.) TURBINL P EVIN U TIIOUT PRESSURIf SPRAY MQ POWER OPERATED IEF VALVES. M.AXIMUM ~~' ??"BACK
; McGUIRE NUCLEAR STATION Finura 15.2.3-8
, . . _ . . ~ . . . . . . . . ~ . . ~ - ,~ , - . - - - . _ - - . - ~ . _ . , _ . . . . . . - - _ _ ~ ~ _ _ . . . .
&h Ck
/
/
4.. 1.
\ ._
3
'd es
- b. .$4 .q
\
as g .. i , 2 . . . 5 9. 640 14 I aed ie3 3g8 fint its
.tl 3
I / l I
*.ti- '
/
- ' c .
J *.Of9 E_ 5 / : ..e.
'd m.
c ..s, l N \A
- o. .
\
N
\
\
. e ,.
se* =ies i.e ,,a i ... itai . it c i- \ MAIN FEEDLINE RUPTURE, McGUIRE NUCLEAR STATION 9 OFFSITE POWER AVAILABLE Figure 15.2.8-1
-3/90 g
k I
- s i
I 8' f k # l D { j t
~'
g .% t g -4. > E I jf, .... tl u I Y- ... l \ 5 I s_ .
.% .i \
N 4 \ 1. gg a ge e , 4 ge3 , te* Ting ggt
/\ t t \
4: *
, \
/
\ \
, q 'x j t
E 5 l 1 %.
- 5ase 3m E
4
=%,
- l l t
)
*tt, l t9 iI 4 ' ieI g3 gg 8 flat stEtt MAIN FEEDLINE RUPTURE, McGUlRE NUCLEAR STATION
@ OFFSITE ::;ure 15.2.8-2 POWER A 3l90
4 v.$4 C.C 'N (48
~
/
EE 48... 4
\
ik 5 g,..... e s t
*.... ' '\N
\
, ,,. w, /i.,- i.'
3,
\
n-x
,n -
y i Ns 1.a
- N f s
\ -
I s ,
, r i
4
/
N./ , in. ' 3
/
~
j s f,....
/
g ....z
.... .! .l \
/
' .. / i..
'iat
.. sci
... ... '\\
\
\
/ \,
MAIN FEEDLINE RUPTURE, rg- FSITE POWER AVAILABLE cGUIRE NUCLEAR STATION
@ Figure 15.2.8-1 3/90
.. . . . . . . , .. .. ~~-....-,--.n.-. . - . . . . - - . . _ . , . _- . . . . . ~ . . . - . .
4 GC D
'*18.
.e.
N 7.... y,,
.Tx i
.o 6... ,
g g.... nat. si! 7 I 13.. il 3~
..t 64, '\ get ,
gg, ettCl
'tt. ,
O. j' A\ y i un un j
= . . . .;
\, .
.k ., ', *
- .... +
CDt.D 1,,
/
.y S...
s...
\to tint 'lICI MAIN FEEDLINE RUPTURE,
',9 OFFSITE McGUIRE NUCLEAR STATION Figure 15.2.8- POWER AVAILAB 3/90
'ff kO R 1 4 i i.... ' i INT ACT STEM q e... Elk (RATCU 3 n. is . < w m . . . . i\ ' 5 g .... , N l E. y .... , s
* \
5 .... , \ rA Tto STIM G
,... . \ ,, tytulot
' \ /
.. gg 8 I i.I q,I
- l. .i in i .. sci
\
'. \
\
1..
! N 1.4 ? s g
I ! ~
\
3 - 3 e I t s N E x
=
... x s s
3 v
.s j __
'.' s.8 i.s i. ,,
11#t it(C6 AIN FEEDLINE RUPTURE. OFFSITE POWER NOT AVAILABLE
- McGUIRE NUCLEAR STATION
- Figure 15.2.8-5 3/90
Yn% ( l l 14t I K I
'- 1 B
W t
- e. t E. . . ,
5 9 g .. t
/
i. se8 ...
\
J t 5 se 7 l i (
- .o r I
.is3*
I I
- ..s.-
1 E E. o .ss
*.sr
...e l
se' i.: i tear esse, l MAIN FEEDLINE RUPTURE, McGUIRE NUCLEAR STATION 9 OFFSITE POtNEfl N Figure 15.2.3 -6 3/90 l
. ~ . . - . ..- - .._. .
. .,+. = - - . . ~ .
. . . . . . ~ . , . ~ - . - - . _ . . . . . _
w YL e 6 4 { 0. .# 8 b g =.6 L,. s g *6. . a. 6 a l y.
,, +4.
\
5 - g !
* =#.1
/
... s i.-
flet 3,,,iltre Il.
-44.
; V 3
- a. j s t . 4 _. ,r
/
7 ( / !. . E. s s. - 5-I [ m 76 - e* o 5-
... . \
t
-se.
se ,as4 143 -14 3 18'
?!nt altCI
-MAIN FEEDLINE RUPTURE, McGUIRE NUCLEAR STATION 9 OFFSITE POWER NOT AVAILABLE Figure 15.2.3-7 3/90 1
i
, .,e.. - ,, -n,. ,.,,
0d 6 2.... l g a.,* Q y r A 44... \ en 2: E so..< y 0 t i ... 7
.. i.i ..a 'a
.. , i.*
tint el(C
/
is et. ,
\-
.ut, -
.n . \
~
- .ei. i ,
g .... t, iset, t g i.... , e ....., \ 2 1- \ g .... t g
..... , \
\
i.s
.e i.
i.
..... '\i.*
MAIN FEEDLINE RUPTU , OFFSITE POWER NOT AV LABLE
, 9 McGutRE Figure 15.2.8 3/90 NUCLEAR
-6 S
. , - . - ~ . . - . ,e.. ..-,.4
- - -- - . - - . . - + - -. _ ,
De tet,
.4 ta Iw.... ; A
= ....
-g COLD.Ltt g ....
440. l##.
- I
%44.
400 len ggd gga, gge tant att a 748.
. . . . I su=n=
N-M w h*' NTI L[G [
!m .... .
s.... _ .. ,/ w ....
.E.
%de. .
I lat. see. se sei sea se8 se' tset attes N\- MAIN FEEDLINE RUPTURE. \ OFFSITE POWER NOT AVAILABLE McGUIRE NUCLEAR STATION l
,. Figurt 15.2.8-9 l 3/90
-. . . - - . . . . - . . . . - . . . . . . - . - - . - - _ ~ . - . . , , ,-
4 __k
~
k i.es. g .ee, liff ACT STEM A GD(ut0AS asse
\- ,
e i....
- - t m
b ***- 7
= l Y *es.
d \ E .se. 'N W
, -l .
tel AULTID STIM
.,ee.
( GDERATOR
- e. - / :
,e, -
i.' i.e N 118 % 4SEC1 N
\ .
i..
\
,!=, b l- h B 'T s
Y
! .. .l
=
-3 E
w ..
-8v I
L ..$ ,e4 ,,, ,,, fing eggg, MAIN FEEDLINE RUPTURE. McGUIRE NUCLEAR STATION
@ OFFSITE POWER N Figure 15.2.8 -10 3/90 New Figure
. _ . - _ _ - _ _ , - - _ - - - . - . . _ - ~ _ _ . - -.
i 1 D& g i. I- t. 5
'y .eo
~
, \\
0 .. i
=
E at
,,, ,,5 i,a ie3 i
flag intti MAIN FEEDLINE RUPTURE, OFFSITE POWER NOT AVAILABLE McGUIRE NUCLEAR STATION Figure 15.2.8-11 3/90 New Figure r
l FSAR SECTION 15.2.3 TURBINE TRIP 120 a 4 l MAXIMUM SECONDARY PRESSURE CASE 100 , =
'[
3 ! 5 t i 80 - 2- !
~ .% '
b - O C '
*1 ;
*$
- f
- l cc 60 th w
- 3: '
O w
- n. We 2 >=4 O
a:
>=
3 40 m > Z' [ + 20 , r t l ! I ^ ^ O 0 10 20. QO 40 7 [ TIME (SECONDS) 4
. ! i ,1 l;f:t ; !i i 4! t:6 le I vl ?f I: : c!f.!IIiI mgc,O w . WeJk 0
4
. ^
P I R T E S A C 0 E E R 4 N I U. S S ' E " B R P R Y R A
^
U D N O T C E, S
)
3 M t S U D 2 M N X I 0 O A 2 C 5 M E S 1 ( E N M I T O I T C _ E S 0 R. _ 1 A. _ S F 0 0 0 0 0 0 0 5 3 1 9 7 5 6 6 6 5 5 5
^a. 0tbo E iG<EE2 t04muN o.OOJ s
i t
;j ,
N ' i i ,
!!. iL j! - ;! !
f.Mc m
. f>*t&
0 4 P I R E T E S A C E R N 0 3 U N I S S E R B P Y R R A D U J f O T C E S 3 M ) J S M D 2 X A 0 N O 2 5 M C E 1 S ( E N M I O I T T _ C _ E S 0 R 1 A S F 0 0 0 0 0 0 1 9 8 7 M 5 a. J$u Wmfg!ga
- C
. . ,i. i
l' FSAR SECTION 15.2.3 TURBINE TRIP 1400 - -r MAXIMJM SECONDARY PRESSURE CASE '
-I i
i
.c 1300 ct 2
E 1200 ,
'N~ .
g Os Z hJ U ' u llE i ! W ti 1100 - /- : f t i 1000 O 10 20 30 ,O i TIME (SECONDS) l l
" [ - , ;; ;! ;! -
<t , ;! . (( i Ya 6 c e ~=.N.W 4 0
4 P I R E T S A C 0 E IE R 3 U N I S S E R B P Y R R A D U f O J T C E S 3 M ) S A f D 2 X
.i A 1
0 N O 2 5 1 M
)
4 C E S ( E N _ M I O I T _ T C _ E _ S 0 R 1 A S F 0 . 0 0 0 0 O 0 0 0 0 - 0 0 0 0 O 0 0 0 0 7 0 5 4 ? 2 1 0 9 2 2 2 2 ? 2 2 2 1
^cTh m 3$mE eeE3gE
FSAR SECTION 15.2.3 TURBINE TRIP- .i 2700 . MAXIMUM PRIMARY PRESSURE CASE 2600 2500
^
0 i en E
~ i 2400 3
n r a C m O E CL 2300 m e a Y m
= 2200 2100 x !
a i 2000 t i 1900 0 10 20 30 40 TIME (SECOND3)
m 2 !
-r FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK ;
120 i. t i 100 a i 't J v 4 A N M 1 x y i 5 J BC ' O z C u. 9 l i O * ! i
>-. t' w 60 kJ "a
u . 00
- a. l uJ I e :,
O 40 i U t i i p ! i l 20 g 1 il 4 l; O 2000 N
- t 0 1000 l!t. ,
l TIME . (SECONDS) ' ll
- .- .. , .. - ,. - - - -- -_- = ._ . ..
i ! FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK
~
=,
630 - b d-iE O l "o 610- {
- E % - -
3Q
- 3 kE k en W
$ 590 -
[ is O 00 f U [a ! O FAULTED LOOP
- I $70 -
----- RITACT LOO 6%
SATURAllOti l yf _ ~ . 1 i 1000 1500 2000 2500 3000 0 500 4 i TIME (SECONDS),
.= .. .- .- .- . ._ _ _ _ _ _ _ _ - _ - - _ - _ - _ - _ _
i a I. FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK ; l
!- 2500
't i 2400 -
- ! [,
, E 2300 --
' ~ -
m 3 M sc C S. W "rz ! 1 , W f i N N
, 2200 - -
w !J QC h b , W l E Lt
}
8L 2100 - I r f 6 N ' ' 3000 1000 1500 2000 2500 l TiesE (SECONDS) ', t r M e e- im T- ,-&ww-e,-+y - +r+
4 i FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK l 90 - - { l 80 - 70 - i / ., J 60 - 5 ' 2 5 1 N ff 50 - 4 L E "~ k i j M-4 10 - 0 - O 500 1000 ' 1500 2000 2500 3000
- - TN8E (SECONDS) e r .m . , . - . .. -- - - < -,__.,c_ _ _ - .
b FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK 1.2 . 4 i 10 , t r 3 FAULTEDLOOP l t-4 3 .......... e n g i m ,3 m O 08- - ? Z UC i ! O- C
, F 9 !
i ' O C j 4 . J E
- w
- g 06- e DJ
- 3. O QC I
d - [en ! 05 2 o l' 3 E l 04-
- O i.
! " I I I i 02- ., < l I i 1 l' k-e u - r 0.0 ... . . . . . . . . . . O 500 1000 1500 2000 2500 3000 l l TmsE (SECONOS) i
FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK 1500 l 4 I h J.r -" n 9 { 1000- - - - - - - - - - - - - - '
- ~ ~ ~~~~ ~ ~ ~ ~ ~ ~ ~ ' ' ' ~
3 4 m - - - g i cr ----- INTACT LOOPS n 5 l 1 l $ FAULTED LOOP d tc a Y tJ m QC Z . 2 m I N 500 - - i I i ,
" l w
1500 <000 2500 3000 0 500 1000 4 TIME (SECONDS)
l l'
- FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK i-3000 l
i
- i. ' ~
! o 2000 - - -- ~- - - - - - - - - - - - - - - - - - - --- - - - - - - - - - - - - - - - q j h' 6 2 ^i E
=
[ 5
=
w O i $ Y
.t J l x .
QC 4 i 5 k tt) -
,000 _ _
p 400 500 600 700 800 900 1000 0 100 200 300 4
- TitBE (SECONDS)
FSAR SECTION.15.2.8 - FEEDWATER SYSTEM PIPE BREAK 12 1.0
- a Z
5 08-Z m m O 2 i
!g 0.6 -
\
4 M ss. oc 3 d o,. .- , t; m. 1 l 0.2 i .i 1 00 - - - 0 2 4 6 8 to 1 TiteE (SECONDP)
. _ _ _ . _ . _ _ _ - . ~ . . . _ . . _ - _- . . - ___ _
- FSAR SECTION 15.2.8 - FEEDWATER SYSTEM PIPE BREAK 2.1 i
i 2.0 g 4 M 1.9 2
- 3Q i tr C to M
C Z O p ( f21 IJ ic i 1.8 - - F i
- 1.7 i
O 1.6 . . 0 1 2 3 4 5 I TIME (SECONDS) i
r 15.3.1.2 Analysis of Effects and Conseocences Method of An& lysis The loss of one pump with four loops in operation has been analyzed. p This transient is analyzed by.t.eee digital computer codes. First, the W IUE.TR AN 02. Code (Reference 1) is used to calculate the loop and core flow during the tran-sient, ir:n; M;;, the and time of reactor the primary trip pressure system based on the and calculated temperature flows, theThe transients. nuclearhe t power,4 VlM E o r :T"*1 Code (Reference 2) H th:n ;eo iv s.Liet, the nem i L tr:n': t i M::d er th: n;':;r ;:1:7 :nd fi; fr;; LOTTR*". Ti nel ly , .o. I:3"C Cede (;;; L o va 4.4) is used to calculate the DNBR during the transient based on the h::t 'b ':; . "'OT"'N :nd 'L i vm LOTT% e The ONBR transients presented represent the minimum of the typical or thimbi cell. Mab sbc.d. Cet e De.s ga M cGe dd gh--* 'MTON 01 hudary c.* add. ions,
% rert Thisaccidentisanalyze#withthe$aracteristicsandinitialconditionsare n a ; " ::::.r;-as described in Reference 3. Plant c discussed in Section 15.0.3.
Initial conditions
'nitial reactor power, pressure, and RCS temperature are assumed to be at
'eir nominal values. Uncertainties in initial conditions are included in the mit DNBR as described in Reference 3.
.'eactivity Coefficients 4 teast. ne=tive. Doppts.t 4.s-patrotom c.oeMiched: is usad,
$ed conservafively large absolute value of the Doppler-only power coefficient iT used (see Figure 15.0.4-1). This is equivalent to a total integrated Doppler reactivity from 0 to 100 percent power of 0.016 .ik/k.
j% most alloweel by ~~a~ hnical h positive mooerator temperature coefficients:f " m " is assumed since this results in the maximum core power during the initial part of the transient when the minimum DNBR is reacned. i Flow Ccastdown calculaiad by RET % N.01 The flow coastde ? ^__ ^4:igis based on a momentum balance around each reactor
- cesiant loop and across the reactor core. This momentum calance is combined
- with the continuity ecuation, a pump momentum balance and the pump characteris-tics and.is baseo on high estimates of system cressure losses. The conserva- -
tism of the calculated flow coastcown is confir?,eo oy comparison with the startup test data of the McGuire units. Results 5 Figures 15.3.1-1 through 15.3.1-# show the transient response for the loss of one reactor coolant pumo with four loops in operation. Figure 15.3.1-/ shows the ONBR to be always greater than the limit value. 5 Since ONB does not occur, the ability of the primary coolant to remove heat from tne fuel rod is not greatly reduced. Thus, the average fuel and clad temperatures oc not increase significantly aoove their respective initial values. l 15.3-2 1986 Update
The calculated sequence of events tablev for the twfcases analyzed shown on Table 15.3.1-1. The affected reactor coolant pump will continue to coast down, and the core flow will reach a new equilibrium value corresponding to the number of pumps still in operation. With the reactnr tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed. 15.3.1.3 Environmental Con:;4uences t . A partial loss of reactor coolant flow from full load would result in a reactor i and turbine trip. Assuming that the condenser is not available, atmospheric ' steam dump may be required. The radiological consequences resulting from ' atmospheric steam dump would be less severe than the steamline break event analyzed in Section 15.1.5 since fuel damage as a result of this transient is not postulated. 1, . 3.1. 4 Conclusions - The analysis shows that the DNBR will not decrease below the limit value at any ? ti:a during the transient. Thus, no fuel or clad damage is predicted, and all applicable acceptance criteria are met. , 45.3.2 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2.'1 Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps. If the reactor is at power at the time of the accident, the immediate effect of loss of coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly. Normal power for the reactor coolant pumps is supplied through buses from a transformer connected to the generator and through the offsite power system. i Each pump is on a separate bus. When a generator trip occurs the buses continue to be supplied-from external power lines and the pumps continue to supply coolant flow to the core. 1 The following signals provide the necessary protection against a complete loss of flow accident:
- 1. Reactor coolant pump power supply undervoltage or underfrequency.
- 2. Low reactor coolant loop flow.
The reactor trip on reactor coolant pump undervoltage is provided to protect against conditions which can cause a loss of voltage to al1 reactor coolant pumps, i.e., station blackout. This function is blocked below approximately 10 percent power (Permissive 7). The reactor trip on reactor-coolant pump underfrequency is provided to trip
-the reactor for an underfrequency condition, resulting from frequency dis-turbances on the power grid. Reference 4 providec analyses of grid frequency _
disturbances and the resulting Nuclear Steam Supply System protection require-L ments which are generally applicable to the McGuire units. l t 15.3-3 1984 Update
The reactor trip on low orimary coolant loop flow is crovicea to protect against loss of flow cenaitions wnien affect only one reactor coolant loop. *his func- ! tion is generated by two out of tnree low flow signals per reactor coolant loop. ; Above Permissive 8, low flew in any loop will actuate a reactor trip. Between . approximately 10 percent power (Permissive 7) and the power level corresponcing I to Permissive-8, low flow in any two loops will actuate a reactor trio. A complete loss of fcreea reactor coolant flow is classified as an ANS condition III event, an infrequent 'ault. See Section 15.0.1 for a discussion of conoition III events. 15.3.2.2 Analysis of Effects and Consecuences The loss of four pumps with four loops in operation has been analyzed. I
~
h A x e_ v 4 L A h k _ .:e This transient is analyzea by tnree digital computer codes. The loop and core- ' flows are calculated by the LOFTRAN code (Reference 1). to calculate the time of reactor trip based on the flows, the nuclear power LOFTRAN is also used [ ! transient, and the primary system pressure and temperature-transients. The ! - FACTRAN Code (Reference 2) is then used to calculate the neat flux transient based on the nuclear powee and flow from LOFTRAN, Finally, the THINC Code (see e
- section 4.4) is used to calculate the ONBR during the transient based on the heat flux from FACTRAN and flow from LOFTRAN. The ONBR transients presented i represent the_ minimum-of the typical or thimble cell. J .
-The method of analysis and the assumptions made regarding initial operating conditions and reactivity coefficients are identical to those discussea in Section 15.3.1, except that following the loss of power supply to all pumos at power, a reactor trip is actuated by either *eactor
- coolant pumo power sup-ply undervoltage or underfrequency. ,
Results 6 Figures 15.3.2-1 througn 15.3.2-f show the transient response for the loss of power to all reactor coolant pumos with four locos in operation. The reactor . is assumed to oe tripped on an undervoltage signal. Figure 15.3.2-/ snows the DNBR to be always greater.than the limit value. 6 Since ONB does not occur, the ability of the primary coolant to remove neat from the fuel rod is not greatly reduced. Thus, the average fuel and clad temoeratures do not increase significantly above their respective initial values.
-The calculated secuence of events is shown on Table 15.3.1-1. The reactor coolant cumps will continue to coast down, ano natural circulation flow will eventually be establishea, as demonstrated in Section 15.2.6. With the reactor tripped, a stable plant condition would be attained. Normal plant snutdown may then oroceed.
15.3.2.3 Envi-onmental Consecuences . A complete loss of reactor coolant flow from full loac results in a reactor and turbine trip. Assuming that the condenser is not available, atmospneric- -{ steam dumo would be reouired. The quantity of steam released would be the 15.3-4 1985 Update
,,v - ,w+wwy,--w---'.------~w-v --,-+r --*w- -y -W -9*
~ . _ _ . _ _ _ . . . . . _ . _ . _ _ _ _ _ _ _ . _ _ . . .
l This transient is analyzed by two digital computer codes. The system ! thermal-hydraulic analysis is performed using RETRAN-02 (Reference 1), l RETRAN-02 calculates the core inlet flow, core inlet temperature, core exit pressure and core average heat flux during the transient. The
.VIPRE-01-code (Reference 2) is then used to calculate the transient -l
-DNBR based on the RETRAN-02 boundary conditions. ;
i f i l I f a I 8 3 p F
. ~ . _ _ - . _ _ _ _ . _ :._______-_ ._ ... _ __._ _. _. _ _ . _ .._ _ ,. _ , _ . . _ . _ ,
i i same as for a loss of offsite power. Since fuel damage is not postulated, the radiological consecuences resulting from atmosoneric steam cump .oula :e less severe tnan tne steamline break analyzed in Section 15.1.5. 15.3.2.4 Conclusions The analysis performed has demonstrated that for the complete loss of forced reactor coolant flow, the ONBR does not decrease below the limit value at any time during the transient. Thus, no fuel or clad damage is predicted. and all
-applicable acceptance criteria are met.
15.3.3 REACTOR COOLANT PUMP SHAFT SEIZURE (LOCKED ROTOR) 15.3.3.1 Identification of Causes and Accident Description ! The accident postulated is an instantaneous seizure of a reactor coolant pump rotor such as is discussed in Section 5.5.1.3.5. Flow through the affected-reactor coolant loop is rapidly reduced, leading to an initiation of a react 1r trip on a low flow signal. Fellowing initiation of the reactor trip, heat stored in the fuel rods conti-nues to be-transferred to the coolant causing the coolant to expand. At the- # same time, heat transfer to the shell side of the steam generators is reduced, ; fir'it because the reduced flow results in a decreased tuce side film coeffi-cient and then because-the reactor coolant in the tubes cools down while the-shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, com-- bined with-reduced heat transfer in the steam generators, causes an insurge into the pressurizer and a pressure increase throughout the Reactor Coolant System. The insurge into the pressurizer compresses the steam volume, actu-ates the automatic spray system, opens the power-operated relief valves, and opens the pressurizer safety valves, in that sequence. The three power-operated reliefvalvesaredesignedfor-reliableoperattonandwouldbeexpectedtofun-ction Droperly during the accident. However. tor conservatism, their pressure reducing effect as well as the pressure reducing effect of the spray is not in-cluded in the analysis. A reactor coolant' pump shaft seizure is classified as an ANS Condition IV ' event, a limiting fault. See Section 15.0.1 for a discussion of Conottion ly L events. I. = 15.~ 3. 3. 2 Analysis of Effects and Consecuences
'The reactor coolant oump shaft seizure transient has been analyzed for one loop seized with four loops in operation. ,(g g g, E I" Method o'f Analysis NIPRE*0\
l RETR % C2 l Two digital computer codes are us d to analyze this transient. The F 9 i Code (Reference 1) is used to ca culate the resulting loop and core' flow ' transients-following the pump s izure, the time-of reactor trip based on the
-loop flow transients, the nucle r power following reactor trip, and the peak pressure, The thermal behavio of the fuel located at the core hot spot is investigatea using _the EMHbm Code (Reference '
2), which uses the core. flowgand W nuclear ::wer :31cd atec :v 2 - - N i ._ . .a, ; .
..v.l.g 47 *=#C- .
- f'ic
- h . W W '02 h
Sed 40% atfackmced 15.3-5
- . - - . - . - . - _ _ . - _ - , - - _ - - _. ~,
l 1 I The result of this VIPRE-01 analysis is a power peaking limit which yields a DilDR equal to the limit value. All fuel pins exceeding this -) peaking limit are assumed to undergo D14B and subsequently fail. l l l k
?
i N i s s P f f f
--_._.--_.,....,.,.-..t..
At the cegirning of the costulatea iockea rotor accident, i . e. At tre time tne snaf t in :ne of the reactor coolant cumos is assumed to sel:e. tne clant is assumea to ce in operation uncer the most acverse steacy state coerating conaition. i.e. , maximum guaranteea steacy state tnermal cower, naximum steacy state cressure, ana maximum steacy state coolant average temperature. Plant characteristics ana initial conditions are furtner aiscussed in Section 15.0.3. ' 60 For the peah pressure evaluation, the initial cressure is conservatively esti-mated as Wesi aeove the nominal pressure of 2250 psia to allow for errors in-the pressurizer pressure measurement anc control channels This is cone to._ obtain the nignest possible rise in the coolant press - during tne transient. The pressure response shown in Figure 15.3.3 the point in tne Reactor-Coolant System naving the maximum pressure. ;2 ie ~- and for o ehr odon to e.sidel'esk aleM - sur r pressuriste bor.m 44% Evaluation of the Pressure TransienQ P Af ter pumo seizure, the neutron flux is rapidly reauced by control roo inser-- tion. 400 motion is assumed to begin one secena af ter the flow in ine affected 1000 reacnea g cercent of nominal flow. No creait is taken for tne cressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlleal f eeawater flow af ter plant trio.
%.6 ;
91tnougn tnese coerations are expected to occur and would result in a lower peak pressure. an additional degree of conservatism is provideo ey ignoring
'" M*4 a.Nes,su n L Mr1% w t.e.k,eyI the, oJyad.ed Mt $d *M WMS War M 4 vt t
P f open at GThe pressuri:er safety valves are tull.foro.13. steam relief is as described in Section ' Evaluation of 'NB in the Core Ouri o the ac ticent For this acci:ent. CNB i: :::r;; u occurs in tne core, and theref ore MIMt44 s an e orehts
- * " - of tne consecuences with rescect to fugi roa thermal transients is per-formea. Jesults cotained from analysis of th3% "not spot" cona1 tion represent the vocer limit with respect to clad temperature,2-- ' c ra d ' - _ a. . - >
l l ? .- + * . . aq'2 <:- -
.+e-e- , ,. , n; g n ; : __ ; - ,;
15: ..e--.g; :: : Se ' ' . a E - 15 3 2t 'u ' #*4d x,m .a. =,;- V ptl bt l [FilmBoilinaCoefficient I The film coiling coefficient is calculatec in :ne F ACTRAN Coce usinc De Bis * ( l hoo-5anceergeong film coiling correlation. ~5e fluia properties are evalua-( tea at film temoerature (average cetween wall anc bulk temoeratures). The program calculates the film coef ficient at every time step basea upon :ne actual- neat transfer conditions at the time. The neutron flux, system cres-sure, culk censity and mass flow rate as a function of time are usea as cro-gra..i incut. _ tt is,
. For tnis analysis, tne initial values of inegoressure m it: J b is.we usea tnrougnout the transient since "r m the most conservative with
. . . - < s m .a . LO -u n o ....o .s respect
,. m . , . tohclaa . w _.a temoerature
_ , .m
. . . . - -^-" \.':
m,,
. ,= , ..
15,3 4 _ _ .._ _ _ __ _ _ _, _u _ __ _ _ _ _
. _ _ _ _ . _ . _ _ . _ _ . . _ __ _ _ _ _ _ _ -._____.____...._.m_
t Oe.Y Fuel Clac Gao Coef ficient _. l The magniture and time depencence of the neat transfer coef ficient between fue and clad (gap coefficient) has a pronounced influence on the thermal results. The larger the value of the gap c: efficient, the more neat is transferred bet-ween cellet and clao. Based on investigations on the effect of the gap co-efficient upon the maximum clad temoerature during the transient, the gap co-ef ficient was assumed to increase from a steady state value consistent with initial fuel temperature to 10,000 BTU /hr-f t 2 'F at the initiation of the tran-sient. Thus, the large amount of energy stored in the fuel because of the small initial value is releaseo to the clad at the initiation of the transient. Zirconium Steam Reaction The :irconium-steam reaction can cecome significant aoove 1800'F (clad tempera-ture). The Baker-Just parabolic rate ecuation shown below is used to define- 1 the este of the zirconium steam reaction. d(w2) = 33.3 x 108 exp -45.500 ot 1.986 7 where:
- w = smount eacted, mg/cm2 t = time, see T = temperature, 'F-The reaction heat is 1510 cal /gm.
The effect of zirconium-steam reaction is included in the calculation of the
" hot soot" clad temperature transient, t
Results gg gggggg hgA % , Drd oc h ted The transient results V are shown in Figures 15.3.3-1 through 15.3.3-/c 1[The . results of these calculations are also summarized in Table 15.3.3-1. The peak Reactor __ Coolant System pressure reached during the transient is less than that wnich .ould cause stresses to exceed the f aulted condition stress limits. Also, the ceak clad surface temoerature is = 's. % --- "0: c It
,a,- -,s- .%, ..m .,o . _ - - _ - - -- . ,v,4. a- - - ~
leili{t.2:a m . L.7'E$$isU ise i$i4C 6oSE Ogwe 15.M-6. !- The calculated secuence of events is shown on Table 15.3.1-1. Figure 15.3.3-1 l shows that the core flow rapidly reatnes a new eouilibrium value. With the ! reactor tripped, a stable plant condition will eventually be attained. Normal plant snutcown-may then proceed. M4*.kment 2. l _ LSert hem /a{conclusions
.5.3.3.
- 1. Since 'tne ceak Reactor Coolant System pressure reacned during any of the transients is less than that which would cause stresses to exceed the fault-e ::-ci tion ". ?s s ' 'mi ts . "? 4-tegri+; P "e dmary ::ci rt 1/ stem is not encangereo.
15.3-7 1984 Update
_ _ _ _ . . - . . _ . - . . _ . ,_... . _ - .._.-__.---_m _ _ . . . . .. _ __ - _ . .- - . - . _ _ _ _ .. . _ _ . _ _ . _ . _ _ . . _ . - . . _ _ _ _ Attachment 1: With the exception of the primary system pressure, the transient results of the locked-rotor peak pressure analysis are virtually identical to those from the DNBR / peak clad temperature calculation. The peak Reactor Coolant System pressure from this analysis is presente_d in Figure 15.3.3-6. ; I b t 0 ) l l
-- , _ - , , ,_, ,~, - . - , - .. ,. . . . . , - .,,,,.wm . .-- . ._... _. ,. . . - . . , , , - - - - . . .- - - _ . _ _ . . . _ ,
l l Attachment 2: 15.3.3.3 Envirnnment al consommnces 1 1 The postulated locked rotor transient causes a reactor and turbine trip, which subsequently results in atmospheric steam durrp. For conservatism, the condensate system is assumed to be inoperable, i thereby limiting the amount of iodine removal prict to at mospheric
- t. team dump. The atmospheric steam ciump does not result in the telease l of any radioactivity unless there is leakage from the RCS t o the secondary system in the steam generators. A conservative analysis of the potential offsite doses resulting frem this accident is pr esent ed considering equilibrium operation based upon one percent defective fuel and a 1.0 gpm steam generator leak rate prior t o t he accident At the time of the accident, 25 percent of the fuel is assumed to fail, thus releasing a significant amount of the gap inventory to the RCF. Dased on this, the following two cases are analyzed:
Case 1: Normal equilibrium Technical Specif i cat ion iodine concentrations exist at the time of the accident (see Table 15.0.9.2). 1. Case 2: I' There is a pre-existing iodine spike at the time the accident occurs. The reactor coolant concentrations are
*he maximum permitted for full power cperation; i.e., 60 times the normal concentrations.
The fol]Lwing assumptions and parameters are used to calculate the activiti release and subsequent offsite doses for the postulated locked ritor accident:
- 1. The total primary-to-secondary leak rate (PSL) is 1.0 gpm. For t hi duration of the accident, this leak rate is evenly dittributed between the four steam generators; i.e., 0.25 gpm PSL in each steam generator.
- 2. Offsite power is lost when the transient occurs.
- 3. All noble gases which leak to the secondary side are released via the steam release without any reduction due to removal mechanisms.
- 4. The iodine partition factor during the accident is 0.01.
- 5. Other assumptions are listed in Table 15.15-2.
Based on the model described in Appendix 15A, the thyroid and whole body doses are calculated at the exclusion area boundary and the low population zone. The results are presented in Table 15.0.7-2. The j doces are within the limits of 10 CFR 100. l
r i MCf e.ays o N g M g Q m M v. Q , Mre, a no % p. on l
- 2. Since the peas. claa urf ace temperature calculated for tne hot spot during >
the worst transient - f u r u i n m ;, a w w J 7GCT e..e av -iii it min 4, e m m 4m+ y. : 9 m ' r d core cooling capability. : 15.3,4- REACTOR COOLANT PUMP SHAFT BREAK 15.3.4.1 Identification of causes ano Accident Description The accident is postulated as an instantaneous failure of a reactor coolant pump shaft, such as discussed.in Section 5.4. Flow througn the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the reactor coolant pump shaft seizure event. Reactor trip is initiated on a low flow signal in the affected loop. Following initiation of the reactor trip, heat stored in the fuel rods continues to be transferred to the t coolant causing the coolant to expand. At the same time, heat transfer to the , shell side of the steam generators is reduced, first because the reduced flow- i results in a decreased tube side film coefficient and then because tne reactor ; coolant in the tuDes cools down while the shell side temperature increases : (turbine steam flow is reducea to zero upon plant trip). The rapid expansion : of the coolant in the reactor core, combined with reduced heat transfer in the '
-steam generators, causes an insurge into the pressurizer and a pressure increase '
throughout the Reactor Coolant System. The insurge into the pressurizer compresses the steam volume, actuates the autumatic spray system, opens the power-operated relief valves, and opens the pressurizer safety valves, in that ' sequence. The three power-operated relief valves are designed for reliable operation and would be expected to function properly during the accident. ' However, for conservatism, their pressure reducing' ef f act as well as the pressure reducing effect of the spray is not included in the analysis. { A reactor coolant puxo shaf t break is classified as an ANS Condition IV event, a limiting fault. See Section 15.0.1 for a discussion of Condition IV events. 15.3.4.2 Conclusions The consecuences of a reactor coolant oumo shaft break are similar to those calculated for the locked rotor incident (see Section 15.3.3). The bounding
.results-for tne locked rotor transients presented in Figures 15.3.3-1 thru 15.3.3-4 and summarized in Table 15.3.3-1 are also applicable to the reactor ;
coolant pump shaf t' break. With a failed shaft, the Impellar could conceivauly
- be free to spin in a reverse direction as opposed to being fixed in position
( as assumed in the locked rotor analysis. However, the net ef fect on core flow isinegligible, resulting in only a slignt decrease in the end point (steady state) core flow. For eoth the snaft_ break and locked rotor incidents, reactor
. trip occurs very _early in the transient. In addition, the locked rotor analysis j.
conservatively assumed that ONB occurs at the beginning of the transient. L l 198' U P d 15.3-8 a- . . . . - . .- . - . - . , _ - . - - - .u- -. - -
REFERtNLti /OR SECTION 15.3 Q q <w w' & At4 r.. L e +- 1
- 1. Burnett, T. W. T, et al. , "'.0FTRAN Code Description", ' CAP 7907-P- A (Propriety) and WCAP-7907-A (Non-Proprietary), April. 1984.
- 2. Hargrove, H. G. , "FACTRAN A Fortran - IV Code for Thermal Transients in a U02 Fue1 Rod", WCAP-7908, June 1972.
- 3. Chelemer, H. , Soman, L. H. , Sharp, D. R. , "Improvea Thermal Design Procedures", WCAP-8567, July 1975.
- 4. Salewin, M. 5. , Merrian, M. M. , Schenkel, H. S. and Van De Walle, D. J. ,
"An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs", WCAP-8424, Revision 1, June 1975.
15.3-9 1985 Update 1
i l A Program for Transient Thermal-Hydraulic 0 CC?d,
- 1. -
EPRI, "RETRAN-02: Analysis of Complex riuid Flow Systems", EPRI NP-195 - Revision 4, November 1988, 2.
" Duke Power Company McGuire and Catawba Nuclear Stations Core DPC-NE-2004Pa , Thermal-Hydraulic Methodology Using VIPRE-01",
Revision 1, February 1990. d 1 1 l
<w -
%.,-i.me--+
,,m. w - , - - . ,,-..,,-y.,w- ,% w,. y w- ,-. , ,,., ,<n_,. - - ,, ,.w.yy--..-y y - - , . ,-,,,,y ,y.-.,y -
7,way 9 ...,y,4_...,w_. .w--.y
Table 15.3.1-1 Time Seouence of Events for Incidents Which Cause in a Decrease in Reactor Coolant System Flow Accident Event Time (sec.) l 1 Partial Loss of Forced Coastdown begins 0.0 Reactor Coolant flow 1.5 Low flow reactor trip M
- l setpoint reached 4.
'5 Rods begin to drop A4 ,
Minimum DNBR occurs 3.6 i l Complete Loss of Forced All operating pumps 0.0 Reactor Coolant Flow lose power and begin coasting down Reactor coo. ant 0. 0 pumo undervoltage ?~ trip point reached Rods begin to drop 1.5 n.ss Minimum DNBR occurs Ja
\.o Reactor Coolant Pump Rotor on one pump locks M Shaft Seizure (Locked Rotor) 4.os Low flow reector trip t-et setpoint reached 2 o5 l Rods begin to drop J.M i h imum isC3 pi va w1. 12 l m
- 2. 2.
Maximum clad MY temperature occurs nied 4to~ bbg O [ l I
*556 upaste l
t J t 3 ( Rotor on one pump locks 1.0 l i i Low flow reactor trip 1.05 setpoint reached l Rods begin to drop 2.05 s 5.2 Maximum RCS pressure occurs , i k t f i f I b 1 i t_ i I h i t 1
+
'we.-,-c.,,- + ... ,..gm..v,..y, , _ ._,, _ , ._, _ _
i Table 15.3.3-2 ; E.arameters for Locked Fotor Dose Analysis l
- 1. Data and assumptions used to estinate radioactive :
sources from postulated accidents.
- a. Power level (MWt) 3565 ,
- b. Percent of fuel defective 1
- c. Steam generator tube leak rate prict to 1 and during accident (gpm)
- d. Offsite power not available
- e. Initial reactor coolant activity Table 15.0.7-2_
- 2. Data and assumptions used to estimate activity released.
a< Iodine partition factor for steam release 0.01
- b. Release from steam generators (1bm) .
~
(0-2 hrs) 500,000 (2-9 hrs) 1,200,000
- 3. Dispersion data :
- a. Distance to exclusion area boundary (m) 762
- b. Distance.to low population zone (m) 8839
- c. X/O at exclusion area boundary (sec/m )
(0-2 hrs) 9.0E-04 3
- d. X/O at low population zone (sec/m )
(0-8 hrs) 8.0E-05
- 4. Dose Data
- a. Method of dose calculations Appendix 15A
- b. Dose conversion assumptions Appendix 15A
- c. Dose Results (Rem) see Table 15.0.7-1 i
l l l _.< _ . _ _ . _ , - - . _ . _ . . _ _ _. _ - - . . ., _ . . _ _ .._._._-,a-_._ . _ . _ _ . . - - . _ _
(Q(O CC, 'N
'N 1.4 s
1% D 1. b /- a .6 e d O
'r
.2 8.
- 4. l. y* 5, 4.\ /5. G. 7' 8' %' 88*
iSCCl
/ \
l.4
)
gI.2 5 i. b
! .s[
f .6 a
\
r l 'S 2
/ 's. 1. 2. 5. 4 5. 6. 7. 8. *. 18.
TINC 60CCI / N FAATIAL LOSS OF FORC C y FIACTOR C001XIT FLO'J \ McGUlRE NUCLEAR STATION Figure 15. 3.1*1 con nona-r
. _ - _ . - . . . _ - - _ _ - . . - - . - . . . - - . . . - - - ~ - = - - - .-._-.. - _.
- - - ~ _
Nf9koc.qL
\
l 1.
,/
* .0 g .8 - l N .
u .4 1.2 i l
.. /
- 8. 2. 5.
- 1. 4 TifC 5.-\s. CC' 7 8/ 9 10.
i
/
Y ! i [ I 2sse. , 1 2580.
/
2400. ll g j*=- L/ R2m.
/
/ ,
a g flee.
'/ \
g- \ s tw.
./ e. 1. 2. 5.= 4 5. S. -7 8. 9 18.
/ tiMC ISCCI J
/
/ PARTIAL LOSS OF FORCED
/ REACTOR COOLANT TLOV w McGUIRE NUCLEAR STATION figure 15.3.1-2 1986 UPDATE
. _ ._. . . _ . . _ _ . . _ . . - . . - _ . _ _ _ _ . . . . . . . . . - . .-. . _ 1 --. .-- _. . _ - . _ _ _ . . .
-. . . .- .- . _ - . - --- . - - - - .-.....- . . - . - -.- ~. _~-. - .
(@ % 1 ' g12
'g / [
/
N WERAGE CHANNEL 8 y .. / ; t d 5 < g -.4 Y
.2
- 8. ;
- 8. 1. 2. 5. 4 -5. 's/ 7 a. 9. Is. i flMC 1
/(b(Cl \
/ \ >
./ \
/ \
l.a
-g1.2 ..- \
g ii. g /
/
'\ \
- g. '
HOT CHANNEL t i 5 d
\ i e fI T/ .2
/
g i ' /
/ ~ .. 0. l. 2. 5. a. 5. 6. 7 s. 9. 18. \
flMC 15CCI
/
./ .
/
/ PARTLU. LOSS OF FORCED REACTOR CCCLANT TLCW McGUIRE NUCLEAR STATION \
= Tigute 15.3.1-3 ,
1986 UPDATE 4 _.._._,... _ _._ .. - . _ ._ a. . s , .. , i...,___i,_-__,__,.__...,-
_ . -_... - .. _ _ __...=. __ ___ _ . . . _ . - _ . _ . . . _ ~ . _ _ . . . _ _ _ - _ . . _
.7
@w
'N
/
2.5 2.4 2.2 I / , j i.. / 1.6 1.4 1.2
-- 8 . - .5
.' 1.5 2. 2.5 5. 5.5 4.5 s
/l.- 4 5.
,' : Titic ISCC3 f.
/
,./
PARTIAL LOSS OF FORCED j : RIACTOR COOLANT FLO'.I
/
McGUIRE NUOLEAR STATIO
/
,Y Figure L5.3.; 1986 UPDATE v '
. . _ - , - - - - . - . ~ _ - - - _ . , . . . . . . .
- EQ %
's
'\
\
N \ /
/ -__
x
\
'N\
i .4 -
\
2 Ns / O I.2 N 5 's hli .
's W \
.8
_8- % ~ d .G U-
=
.4 5-
.2 8.
/ \
- 8. 1. 2./ '. 5 4 5. 6. 7. S. 9. is.
/ TINC ISCCI \,
/ \
/ \
/
/ N, .
/ ,
. COMPLTE LOSS OF FORCD REACTOR COOLANT FLOW McGUIRE NUCLEAR STATION Figura 15.3.2-1 i-
./
/.
1986 UPDATE-r
N (h GCD
\ 1.4
\
N2 I - g 1. s
\
E \
"8 .
'\
~
y .6
/
2 I .4 2 4~
.2 8.
- 8. 1, 2. 5. 4K 5. 6. 7 8. 9. L8.
't ISCCI
\ /
\
/
'N 2696.
S 2588. ' U \ 2460. l 1 y2588.
/
- \
$2288. / s-5 !
2188. -
/
2888. ,/
/
- 7. 8. 9 \
- 8. 1. 2. 5. 4. 5. 6. II.
s '
- TINC 45CCI
/ \
\
j/ , j COMPLETE LCSS OF 70RbC REACTOR COOLMIT TLOW l
/ M McGulRE NUCLEAR STATIOf4 W"'8 "" Figure !5.3.0-2
\
l 1986 UPDATE i
%\ \ )
\ '
\ .
'*d Ni /
.s /-
~
i.2 l
\,
/
$ s'h '
h AVERAGE CHANNEL W .8 N\ E \ y e$ S w
- 'N\
.2 g
\ -
- 8. ~
- 8. 1. 2. 5. 4. \ 5. 4 7 f.t i. 18.
flMC ISCCI
\ ,
\
\ /
\ /
'N s
1.4 g8.2 ,/ 1 Y 1.
\ s i- '
- b j HOT CHANNEL \
l V .s E /
/ \ \
,/ . N 3 , '
\
\
i a .. / y . Y s i- \ i
,. 2
/
.g*-!
Nx
- a. l. 2. 5. 4 5. 6. 7. 8.. 9 it.
N l
/,
T ilt ISCCI \
/
\
/.
l
/ CCMPLrrE LOSS 0F FORCID
'f REACTOR COOLA.W f10W s
McGUIRE -NUCLEAR STATION Figure 15.3.2-3
-1986 UPDATE
. __ _ _ a. . . . . . _ . _
. . . .,____ ~...._ . . . . - . - _ _ _ . _ _ . _ - _ . - . . . _ _ _ _ _ _ . _ _ . _ , _ _ _ _ . _ . . . _ . . . . . . . . . _ . -
x Q
'N h
2.8 N . , L N
\\
2.4 \ 2.3
- t. s . .;
l"
!2. /
l/ \ 1.e t l !? 1.s , i
.l.4 j l .2 ,
/ 'N
- 4. .5 - 1. 1.5 .2. 2.5 5. 5.5 4 4.5 S.
ilPC 1. ISCC)
\\~
1 l
.\
! N I
. --/
,- \
i-l COMPLETE LCSS OF FORCED '= REACTOR CCOLA.NT Ti.0W
. McGUIRE NUCLEAR STATION .-
Figure 15.3.2-4 1986 UPDATE
- - - _ - . . - - _ . . - - _ _ - . . . _ -x
1
-. - -1.2000 -
M- R (n'te,,, i
~ 't.0000 - -
E l C
, , 80000 - -
=
N y T0000 < - 3 3 .40000v - - d 0 m .20000 - - d E 0.0 e 1a i J 1 J i a a S TIMy (stt) 1.2000 _ g 1. 0000 <
/ \ ..
5-
* .75000 < '
-6 o- ,
M m 4 50000 - -
-g i' E .25000 <
~
/
l c 0.0 /
-S 23000'<
[- - 9 / 5 l 5 /- 40000 e g o
-g- g;x '
-s
'/ o_
a
-j e n. J & 4
- T!HE ($CC) x N
/. .
/
./- 1.0CKID ROTOR McGUIRE NUCLEAR STATION Tigure-15.3.3-1 1986 UPDATE s ,e,'-- '.,w...,--*.,-,,.,e - . . - . . , g c - . . - w ,,.s .
y
ktplaco
\
\ '
2700.0 N - 2600.0 - -
'N\ / --
'N 2500.0 - - --
- h. _ 2 4 00. 0 -
\ ---
M a-R 2300.0 - - -- 1 N
~
0
- 2200.0 - -
\. --
xx 2100.0 - - -- N
\
2000.0 o
/ .-
/-
. ~
. k e.
8 e k
=x x
=
o-
. -/
/ ':ME (SECC.1DS'
/
/
,/
/
t' o Y l-
@ MUIRk~NDCLEAR i gure 15 3. 3 ;987 decate S
l I
.- -..;. _ . , _ . . ~ . . . . . - . ~ . - .
. . _ . . . - . . . - . - . ~ - -.-
2 ,
^t noea .
'. ' C-e ,
- +- .
f
- 1:03 + +
.l. ! t'.
3 : n:o+ + W- ; i
;g -- . , ,
- 9 :::o +
! i a s -
's *tt:0 v +
3-- \i \ l
- s- x +
= ;
m i!0C0 \ ' to tN ; 1 1 1
. N.i .. . . !. .
\ *:wt tci t 2000
\ /
/' ;
l- .
- :cco *
- i 1 !
j i
- cco + .\ #4 CHANNEL 3 k' j 5:f.00 +
7
/ i I
5
*:000 -
\.
y. 2 :co - - t
- 3. 3 -'
\ $
a ~ . a . - , -s
? wt sto \
- \
2000 .
/ \.
- Ocjr - +-
O /
} /
3
*/
becot
- AVERAGE --::-!NNEL i
?# I
+
l --. ,/ .4::cc + 1.. ./ J -i i:cco + -+- l' t _
- c &. ,
t , 1 i t
^
o 1 N 1 e i 4 i a o
~
*:wt tto
[- ~ LOCKD - ROTOR I ~: L -ms - McGUIRE NUCLEAR STATION Y Tipre 53.3.3-3 1986. UPDATE
.. ._ . . _ . _ _ . . _ ~ . . . . . . . . . . . .__ . . _ _ . . _ . . . _ . ..- . _ ..
/
[.Qlncr. x-
.. s
-\y
[
/
/
/
/
/
3000.0 '
'N - /
N /
'N
'N 2500.0 - -
'N --
t
/
~
s z 2000.0 - - -- w s\ o
/
5 ..
- N -
=
1 -1500.0 - -
/
1000.00 - .
/
/- Nx '
500.00
/
8 o -o 8 8 o 8N o N 8 o
- e. r/ o o. o. -
o x s O y o' N e w a ~
'N T1HE (SEC)
/
f.
/
/
LOCKD RC*0R
* >Y+a8 =
- McGUIRE NUCLEAR STATION Tigure .3.3.3-a 1986 UPDATE I
FSAR SECT::ON 15.3.1 - PARTIAL MOSS OF FLOW s.2
~-~~-~~~ ~~------ ---~~~~~
-------- ~~- ----
1.e 3 _ [ ' faulted loop b - ntact loop 3 7: x _ g e.G c, r G S e.4
~'
- x. y 5 E
[>
\ '
g e.2 U
- x e.e N
-41 . 2 , , , , , , ,
2 3 4 5 6 7 e e le e 1 TIE (ED30SI
, .: 1 ' t i ' ,
- T Tre S ; N 7 u W e O
l L F 9 F O S 8 S O _ L 7 L A ~ ' I ~ T 6 R I A I S P h 5 O B E E 5 E
-. (
1 N 4 E D T 3 . 5 5 - 3 1 E N e, O I 2 T 3 C 1 E I S y R . e A .
. 8" #
S F i2 o 5 58' z ";; z
A
- i'SA:1.SEC"".ON 15.3.1 - PAR"IAL LOSS OF FLOW- ,
l '. 2 3 I ~ e i i.e ~
! N 2 e.a N 3 e - e.s x %
o u, s!
- 9.4 !
o T
~
e.2 ~ N 8 . e.e , , , , . . . . . . . 5 6 7 8 9 IS e i 2 3 4 TIE (EEXBSSI t
TSAR SEC"':0N 15.3.1 - 7ATIAL LOSS 0? Fm0W 2488 2,s. L o
- 23ee m
a - u
- d N
$" - - G i
N 4
. t F
~_
er2 220e t/1
- Lsl i
2850
~
[ zies , , , , , , , , , i 2 3 4 s s 7 e a se l e Tite (FIIsest
i
?SAR SECE'10N 15.3.1 3 ART:A m .o OSS OF 7 LOW ,
2.4 , l 2.3 2 g
= _ _ -
- 2.2 G.
g
\ b N ~~
2.,
\ _
/ .
- 2. 0 ' ' .
I g- G y 8 1 3 4 TifE (MEX3861 , t i r
--a
. FSAR SECTION 15.3.2 COMPLETE LOSS OF FLOW ,
l.0 v e..
~ ,
3 e.a N 3 5 N. ! e.7 N N x 5 x m ' N C e.B x h-o_ * , n ye5 u m p . 3 54 e e -
, e.3 3
o i e.2
-~
e.: I e.e ~ - - i s s 7 e o to j e i 2 3 4 T!fE (EEDSS) , l i i 4 -
1 L l FSAR SECTION 15.3.2 - COMPLETE LOSS 'O.F FLOW 129 ,
~'
ISO n Y
!3 ** m 6c -
\ $
e se ~ g cr W y tu la r 49
> N w
x _ 29 n a u 5 a e s a a s a g 4 5 6 7 8 9 18 9 1 2 3 , THE (EtXBCSI l 1 1
- .- . . . . . . - . . . . . - _ - . . . ~ - .- . _-
Figure 15.3.2-3 y C A . O . En ~ CD
'O .
A " f.zQ - / E- . f=3 A . CL. ~ 2 O / O . N ..
. ,--A I
N
- O O- -
f.il ~ ! Cn
-LY .
.- =
! < ~.: i e i
~
i e Cn h (1VNIMON 30 NOI10TbJ) XOld IT3H 30Td34Y 3 MOD
l 4 t
;! i : 'r,' ,:I -.! ' 'l!in :.i- ' ;I! :
- n
% ng sO ~U.W.tJL e
W e O l L 1 F F 9 O - S 8 S O - L 7 E T . E L N s P . M O
% 5 t
s e D C , E E ( E
- ~
4 T
~
2 . 3 . - 3 5 1 , N 2 O I T - - i C 1 E S R ~ O A 8 9 e. e n m M B W e 0 S 3 6 2 S l G. o 5 M
' S m e 5' 4
5 3 5 P _' Omn e wma*4w"u.5w* 2 Wo4ew>< r uoOa..
,
- i'
FSAR SECTION 15.3.2 - COMPLETE LOSS.0F FLOW 2s 2488 n b h
=
/ N \ 5
! W toa m 2200 N i,a g
" Un E
5 "a 2199
~ i 2eee , , , , , , ,
4 5 6 7 8 9 18 4 e 1 2 3 iIFE (SEIDOS}
FSAR SECTION 15.3.2 - COMPLETE m0SS OF FLOW 2.4 2.2 g x b 2.0 ^ E
$ E'
\ N t,J t.8 1.6 N m .
/
1.4 = ' , = 5 7 O 2 3 4 \ na tSEDROS3 l
! :< ' i t .il .!: lt
[: n 3 I::..- O W U YY ~ 0
. 1 N 9 R p o
lo p o O T O d e lt f lo u! t a a i t c n N 8 R D E f i N 7 K. C O 6 L ) S x D N 3 _ 5 O C 3 E S ( 5 E _ 1 M I - N 4 T N O I T - 3 C x. E S R 2 - A S F [ - 8 6 4 2 0 2 4 2 0 0-1 0 0 0 0 0-2<P o O5O E nog a
I' 'l e FSAR' SECTION 15.3.3'- LOCKED ROTOR 120 100 2
' Z 80 = ,
ci 6e . d y So 5 - U8 3 2 P z U e I
$ 40 E k x
~ x i 20 ; i t I O - 0 1 2 3 4 5 6 7 8 9 10 TIME (SECONDS) i
_za FSAR:SECTION 15.3.3 - LOCKED ROTOR - , 1.1 1.0 - m j 0.9
\
.-ga O
- 0.8 u.
O z o . 0.7 P T e i E ' h S O6 9 - t f'D x o w i LL 5
. N P -;
W I G4 0 4 . 2 LLE \
$ 0.3 r i W
- f g
O O O2 4 I 0.1 i ~~
~
~ * '
O i 2 a 4 "' s * *'"i s 7'~" # a 9 to TIME (SECONDS) , 4;
d FSAR SECTION 15.3.3 - LOCKED ROTOR 630 620 - faulted loop { intact loops 610 { c c 600 m - E .- m % C T f 590 % e e m * , ct fD 3 580 *
~
m
?
W i O Y , @ 570 $ 4 <
- ! ct
@ 560 i
.2 i
550 I. 540 i 530 . , , O 1 2 3 4 5 6 7 8 9 10 TIME (SECONDS)
FSAR SECTION 15.3.3 -.LOCKEDLROTOR 2500 2400 O
$ 2300 I ). ,
S E a E J g
!g 2200 Y -
O
$ 1 i
t 2100 l' 2000 0 1 2 3 4 5 6 7 8 9 10 TIME (SECONDS)
FSAR SECTION 15.3.3 - LOCKED ROTOR 2600 2sw y
/ N y 3
S is
=
2400 , g-
) N 2
a / .ta 2 1 2300 0 1 2 3 4 5 6 7 8 9 10 TIME (SECONDS)
4 i
- FSAR SECTION 15.3.3 - LOCKED ROTOR i 2MO 1
4 4 1 i j , i j i i N _/ h f I I ' ' c^ B s % ' m C c:- J M
> f 7
$ r % >
w tJi
,/
i- E / Y ) cn - Y
- a.
- .?
I
.2300 ,
f i-i 4 2200 . . . . .
' # '5 7I#
i '" 0 2 .'8' 1 3'~ -4 ~ I5 4 "i"'""6' .,9 :: '10 TIME -(SECONDS) 4
- - ____.i__' ^J..__i___-
6 FSAR SECTION 15.3.3 LOCKED ROTOR I 760 t N I 2 m-A D q
~
O trc EQ C z- as - zw ru [$
< t-750 - H Os ,
62 d u> a a 2 ~ I 740 . O 1 2 3 4 TIME (SECONDS) e.
l ( I An uncontrolled RCCA bank withdrawal from a suberitical or low Dower startup conoition is classified as an ANS Condition 11 event, a fault of moderate frequency. Sec 5ection 15.0.1 for a aiscussion of Conditicn 11 events. 15.4.1.2 Analysis of Effects and consecuences Method of Analysis Thektobcs. analysiswrtA of the '%Uncontrolled se ri A" RCCA bank withdrawal from subtritical ; accident is performed in three bsages: first an average core nuclear power' ; transient calculation is perforned, followed by an average core heat transfer I calculation, and finally a DNBR calculation. The average core nuclear calcula-tion is performed using a spatial neutron kinetics code TWINKLE (Refer 3nce 1) to determine the average power generation with time including the various total ccre feeoback effects, i.e., Doppler reactivity and mocerator reactivity. The average heat flux and tenperature transients are determined by performing The a fuel rod transient heat transfer calculation in FACTRAN (Reference 2). average heat flux is next used in THINC (described in Section 4.4) for the+ " transient DNBR calculation. 01 ant characteristics and initial conditions for this accident are discussed in Section 15.0.3. In order to give conservative results for a startuo accident, tne following assumptions are made: .
- 1. Since the magnitude of the power peak reached during the initial part of the transient for any given rate of reactivity insertion is Strongly de-pendent on the Doppler coefficient, conservatively low values as a func-tion of ;ea:r . . w.cu. :.. ecti:r 15.0 ; mov Teb;. 1;.G.;-4 -
' Tnteofetthe6"moderator reactivity coefficient is negligible during Contribution 2.
the initial part of the transient because the heat transfer time between the fuel and the moderator is much longer than the neutron flux response time. However, af ter the initial neutron flux peak, the succeeding rateA of power increase is affecteo by '.he moderator reactivity coefficient. 2261f conservative value i:a ...u in Uiw anoly;i; :: yicid the Taxi- -
+ :.L heat " =. " LM c The reactor is assumed to be at hot zero power. This assumption is more 3.
conserv6*ive than that of a lower initial system temperature. The higher initiai syster temperature yields a large fuel-water heat transfer co-efficient, larger specific heats, and a less negative (smaller absolute magnitude) Doppler coefficient, all of which tend to reauce the Doppler feedback effect thereby incrsasing the neutron flux peak. The initial effective multiplication factor is assumed to be 1.0 since this results in the worst nuclear power transient.
- 4. Reactor trip is assumed to be initiated by power range high neutron flux (low setting). The most adverse combination of instrumant and setpoint errors, as well as delays for trip signal actuation and rod cluster con- i; trol assembly release, is taken into account. 4 1; guis n. .sium..
u$n %rt *esm.m a fe; u. Le .; ..a r fl m + .,;tpu nt ni;i n;; M 'a .u. uvm - C)" n:1 .el i c-f ^i voiwoim .s ;; ui u ut. Since the rise in the neutron flux is so rapid, the effect of errors in the trip setpoint on the actual 15.4-3 1055 spsets
__.___.m___m . - - - _ . _ _ . -. . _ - . _ . . -._. - . _ _ _ ..__ _ _ _ _ _ .-. _ 4 1 l , Insor _L, The analysis of the uncontrolled RCCA bank withdrawal from zero power ' consists of a RETRAN-D2 (Reference 1) plant transient simulation and a ' VIPRE-01 (Reference 6) core thermal-hydraull.- analysis using the Statistical Core Design Methodology (Reference 8). The reactor power ; transient initiated by the uncontrolled red withdrawal is modeled ; using point kinetics. Moderator and Doppler feedback and control rod insertion following reactor trip are modeled. Primary and secondary systems are modeled in detail. Boundary conditions from the RETRAN analysis are input to VlPRE to determine the Jetailed thermal-hydraulic response of the reactor, including the hot function. The ! minimum transient DNBR is determined using the BWCMV and W-3S CHT ' correlation. T n orrt th , fuel average temperature are used. At zero power a Doppler coefficient of -1.325 pcm/*r is used. This value decreases linearly to a value of -1.04 pcm/*r at full power. l Insert C equivalent to the Technical Specification MTC vs. power level ; relationship is used. Between zero and 70% power, t.he MTC is +7 pcm/*F. This value decreases linearly to zero at 100% power. The MTC is modeled as a density coefficient based on the core average moderator temperature as a function of power level. Insert D r The r.ominal high power range flux low setpoint is 25% FP. Due to the , effect of control rod motion on the excore flux signal, a
- conservatively high setpoint of 111.1% 13 used in the analysis.
P I l l l i
The., worth Wdhdrown durq Oe, dronsient is ras rne,rd durmg 44. reati.or tr*p. time at wh'ch the rods are released is negligible, in addition, the re-actor trip insertion characteristic is based on the assumption that the highest w rth rod clutter control assembly is stuck in its fully withdrawn position. See Section 15.0.5 for red cluster control assembly insertion characteristics.
- 5. The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two se-quential control banks having the greatest combined worth at maximum speed (45 inches / minute). Control rod drive mechanism design is discussed in Section 4.6.
- 6. The most limiting axial and radial power shapes, associated with having the two highest combined worth banks in their high worth position, are assumed in the ONB analysis.
- 7. The initial power level is assymed to be below the power level expected for any shutdown condition (10 8 of nominal pow ). This combination of highest reactivity insertion rate and lowest ini;ial power produces the highest peak heat flux.
Three,
- 8. .Iwer reactor coolant pumps are assumed to be it. operation. This is conser-
*hur% Evativg with respect to DNB.
Re s ul ti, ijbWcA. w rk "h5e rt Y " Figures 15.4.1-1 through 15.4.1-3 show the transient behavior for the uncon-trolled RCCA bank withdrawal incident, with the accident terminated by reactor-trip at 35 percent of nominal power. The reactivity insertion rate used is greater than that calculated for the two highest worth sequential control banks, bnth assumed to be in their highest incremental worth region. Figure 15.4.1-1 shows the neutron flux transient. The energy release and the fuel temperature increases are relativtly small. The thermal flux response, of interest for DNB r.onsiderations, is shown on Figure 15. 4.1-2. The beneficial effect of the inherent thermal lag i' the fuel is evidenced by a peak heat flux much less ttan the full power nominal value. There is a large margin to DNB during the transient since the rod surfa:e heat flux remains below the design value, and there is a high degree of suNooling at all times in the core. Figure 15.4.1-3 shows the response of the average fuel end claading temperatures. The average fuel temperature increases to a value lower than the nominal full power value. The minimum DNBR at all times remains above the limiting value. The calculated sequence of events for this accident is shown on Table 15.4.1-1. With the reactor tripped, the plant returns to a stable condition. The plant may subsequently be cooled down further by following ncrmal pla.it shutdown [ procedures. l 15.4-4 N @ i* l
_ . _ - . - - . - , _ _ _ _ - _ _ - . - _ . - . . - . . ~ . - - . i 1 e i I Insert F. ; I
- 9. The steam generator secondary is modeled as a single control -
volume and uses.the RETRAN-02 local conditions heat transfer j option. This approach is consistent with the zero power initial [ condition. , l
- Insert r e
rigures 15.4.1-1 through 15.4.1-6 show the transient response { following an uncontrolled bank withdrawal from 10E-9 rated power. The . sequence of events is given in Table 15.4.1-1. The total reactivity is shown in Figure 15.4 1-1. The reactor goes prortpt critical before Doppler feedback terminates the power excursion. Rod insertion on reactor trip ensures the' shutdown of the reactor. The neutron power transient is shown in Figure 15.4.1-2 and the resuAting heat flux in rigure 15.4.1-3. The heat flux is the key result with respect to , determining the approach to DNB. The fuel average temperature and the .[ core outlet temperature are shown in Figures 15.4.1-4 and 15.4.1-5, respectively. These temperatures are well below the nominal full ; power values. The minimum transient DNBR is shown in Figure 15.4.1-6. The result is greater than the statistical core design (SCD) limit of 1.55 whan using the BWCMV correlation.- For bottom-peaked axial power ; shapes which may approach DNB below the mixing vane grids, the W sS ; correlation is used. The minimum DNDR with the W-3S correlation is significantly greater than 1.3. , a l i . i
-h i
r
-v.~wr,-.- ...,--,w-m-m...we,v,,.,.,-_w-,ww,.-,,,,y.,,,,,,,.,m.y---,. ..m.y,~,--,- ,v.-e,..~,..,,w,,~,..,,.,,,w+
15.4.1.3 Envirenmental cansecuences There will te 90 raaiological 0;nsecuences associatec altn in uncontrailea rac cluster assemoly cank itharawal from a suocrit1 cal or low ocwer start so concition event since radioactivity is containea witnin ta fuel reos anc Reactor Coolant System witnin cesian limits. 15.4.1.4 Conclusions In the event of a RCCA withdrawal accident from the succritical conoition, the core anc the Reactor Coolant System are not aaversely affected, since the com-- oination of thermal oower and the c:olant temperature result in a CNBR greater than the limit value. Thus, no fusi or claa camage is predicted as a result of DNB. 15.4.2 UNCONTROLLED R00 CLUSTER CCNTROL ASSEMBLY BANK WITH0RAWAL AT POWER 15.4.2.1 Identification of Causes and Accident Descriotion Uncontrollea roa cluster control assemoly (RCCA) Dank witharawal at Dower re-sults in an increase in the core heat flux. Since the heat extraction from the-steam generator lags behind the core power generation until the steam gon-erator cressure reaches the relief or safety valve setpoint, there 1: a net increase in the reactor coolant temperature. Unless terminatea by manual or automatic action, the power mismaten and resultant coolant temperature rise could eventually result in DNB. Therefore, in order to avert damage to the fuel clad the Reactor Protection System is designed to terminate any sucn transient before the ONBR falls below the limit value. The automatic features of the Reactor Protection System which prevent core camage following the postulatec accioent ingluce the following:
- 1. Reacter trip 1s actuated if any two out of f0ur Dower range reutron flux instrumentation channels e.rceea an overpower setoolnt.
- 2. Reactor trip is actuateo if an> two out of four AT channels exceed an overtemperature $7 setpoint. This setooint is automatically varten alth axial power 1moalance, coolant temperature and cressure to protect against "NB'
( coo \ent kepva % rt and coolenF kmpen% increase nt it L Reactor trip 1s actuated if any two out at ! cur ai cnannels esceeo 3n overpower 2T setooint. This setooint i s autcmatically ,ar190 i tn =*== o ensure t .at tne al'owaole neat genertt :n ute , a3 15 not exceecea - 4 A reactor trip is actuatea i' 3ny two out of f our cressari:er - assure
- annels e<ceec s fixea setoolnt. ~n is set nressure is .ess : sn tne set oressure for tne cressuri:er safety valves.
- 5. A reactor trip is actuatea if any two out of three cressurizer 'evel cnannels excena a fixea setooint onen tne reactor onwer .s acove sooroximately 10 percent (Permissive- 7).
15.4-5 "4" c 0,
_ _ _ - . . __._._._ ___ _-_._ _._ _ _ _ ____ _.__.____ __m -_ i In addition to the above listed reactor tries, there are the following RCCA witharawal Olocks: , i
- 1. Hign neutron flux (one out of four power range)
- 2. Overpower AT (two out of four)
- 3. Overtemoerature AT (two out of f our)
The manner in which the combination of overpower and overtemperature AT tries provide protection over the full range of Reactor Coolant System conditions is , described in Chapter 7. Figure 15.0.3-1 presents allowaole reactor coolant , loop average temperature and aT for the design power cistribution and flow as a function of primary coolant pressure. The boundaries of operation defined by the overpower AT trip and the overtemperature ai trip are represented as
" protection lines" on this diagram. The protection lines are drawn to include all adverse instrumentation and setpoint errors so that under nominal '
conditions trip would occur well within the area bounced by these lines. The utility of this ciagram is in- the f act that the limit imposed by any given DNBR can be represented as a line. The ONB lines represent the locus of conc 1ttens for which the ONBR equals the limit value.(1.27 f e. ..m m;.el: ;d ' ex 1. M o r the w icel cell). All points below ano to the left ofThe a ONB line for a diagram shows given pressure have a ONBR greater than the limit value. , that ONB is prevented for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable ONBR line et any point. The area of Dermissible operation (power, pressure ano temperature) is bounceo by the comoination of reactor trips; high neutron flux (fixed setpoint), nign pressure (fixeo setpoint), low pressure (fixea setpoint), and overpower ana overtemperature AT (variaole _setpoints).
'The' uncontrolled RCCA bank withdrawal f rom full power is classified as an MS Condition II event. a fault of moderate frequency. See Section 15.0.1 for 3 discussion cf Concition II events. ,
15.4.2.2 Analvsis of Effects and Consecuences T Method of Analysis RFrRAN-02 } The transient is analyzec by the C"""' Code (Ref erence /). This coce
' simulates tne neutron kinetics, Reactor Coolant System, pressuri:er, pressuri:er *elief ano safety valves, pressuri:er scray, steam generatcr. vc .
steam generator safety valves. The.coce computes certinent plant variacles including 'emoerature, pressures, and power level. Sc :on -it: n
intrates - " Qure it n 1-1 tra m ec n 'ncut : .:F"" t: ceter-i m tie-
-+a :T q tv =%
andh MU*MCO Mudhal Co re ' ^ This accicent is ene4444m ( Tho DHSR calculaNed for ph,nt
-. e the 6mc ":t Design Procedure ^as ,
descricea in Referenc'e d clant cnaracteristics and initial conditions are discussed in section 15.0. In oraer to cotain conservative results for an uncontrolled rod withdrawal at puwer accider*. the following assumptions are s mace:
$ s O & ,(taft.c N Of.
i 15.4-6 l --. .- - . . - . -
- ~.- . - - , - . , - , - . _ - _ _ _ _
- l. Initial reactor power, pressure, and RCS temperatures are assumed to be at their nominal values. Uncertainties in initial conditions are included in the limit DNBR as describea in Reference 3.
Stokota. w d.k fWl.ac.Le= A l
- 2. Reactivity Coefficients - Two cases are analyzed:
- a. Minimum Reactivity Feedback. A least positive moderator density f coefficient of reactivity is assumed corresponding to the beginning of core life. A variable Doppler power coef ficient with core power is used in the analysis. A conservatively small (in absolute magni-tude) value is assumed.
- b. Maximum Reactivity feedback. A conservatively large positive '
moderator density coefficient and a large (in absolute magnitude) negative Doooler newer coef ficient are assumed. 11 3 . 2. m
- 3. Thereactortriponhigh]neutronfluxisassumedtobeactuatedata conservative value of .13 percent of nominal full power. 3 The ai trips include all adverse instrumentation and setpoint errors. The delays for trip. actuation are assumed to be the maximum values, i ne, c.alcuMad neidrog ,
flux ts constevetavelg mod And to account for tronsie,3 C.%ts in he. h irdJant c-
- 4. The RCCA trip insertion characteristic is based on the assumption that the 94 highest worth assembly is stuck in its fully withdrawn position.
((Q
- 5. The maximum positive reactivity insertion rate is greater than that for the simultaneous withdrawal of the combinations of the two control banks having the maximum combined worth at maximum speed.
The effect of RCCA movement on the axial cora power t 'tribution is accounted for by causing a decrease in overtemperature'AT trip u point proportional to a decrease in margin to DNB, as 4t.5Cratd in Rt:9tsence. (2.. A discussion of ATWT considerations is presented in Reference 5. Results t Daw - Figures 15.4.2-1 through 15.4.2-3 show the transient response for a rapid RCCA withdrawal incident starting from full power. Reactor trip on high neutron flux occurs shortly after the start of the accident. Since this is rapid with respect to the thermal time constants of the plant, only small changes in T g9 uand press _ure result and maroin to ONB is maintained. ( 5t%nWitont 10 % The tra ient response for a slow RCCA withdrawal from .f.M t power is shown in Figures 15.4.2-/lthrough 15.4.2-E!4 Reactor trip on overtemperature AT occurs aftcr ai;re , m mm -u - rise in temperature and pressure. h ::"^curM'y hr;r tP 'r n;id CA iedrrd. % The minimum DNSR is greater than the limit value, g g 5 / Figure 15.4.2-7 shows heminimumDNBRasafunctionofreactivityinsertion rate from initial apower geration for i i r R - 9 9 reactivity feed-back. It can be seen that two TYactor trip #M provide protection over the whole range of reactivity insertion rates. These are the high neutron flux;Wah , and overtemperature AT t .,,13. The minimum DNBR is never less than the limit oressuti value. Med%s N rt 15.4-7 6 Whw
I 2.. Reactivity Coefficients - Two cases are analyzedt
- a. The most positive moderator temperature coefficient (MTC) allowed by the Technical Specifications is implementedA as a least positive moderator density coefficient (MDC). I Doppler temperature coefficient (DTC)Avariable conservatively with fuelsmall I ternperature is used in the analysis. r' (in absolute magnit ude) value, which corresponds to beginning of core life (BOL), is assumed.
b. A conservatively positive MTC, corresponding to end ofA core DTC i life (EOL) , is-implemented as a least positive MDC. A conservatively variable with fuel temperature is used.which corresponds to end small (in absolute magnitude) DTC, of core life, is assumed. I P i
?
I 9
~. , r ~ ....,..~..y ,...,y...,_- . - , . . w ,,,w.,.,-m., .-,...,,..--w..r -m... __,.._,,r_,,,, v..e.. ,.. , , .. ., . , , , ..
- _ , _ _ ._- . - - ._ - ~~ _-_- ,
6 7 Figures f" reactivity15.4.2-gand15.4.2-gshowstheminimumDNBRasafunctionoMand100 insertion rate for RCCA withdrawai incidents starting at percent power, respectively. The results arr similar to the 10# percent power case, except that, as the initial power is Mcreased, the range over wnich the overtemperature AT trip is ef fective is greased. In neither case does the DNBR fall below the limit value. The shape of the curves of minimum DNB ratio versus reactivity insertion rate in the referenced figures is due both to reactor core and coolant system tran-sient re'ponse and to protection system action in initiating a reactor trip. Referring to Figure 15.4.2- for example, it is noted that: 7 90L
- 1. For high, reactivity / insertion rates-(i.e. b:tu: = 5 a 10 iissec anu 1.0 x 10 ' it'n:)/ reactor trip is initiated by the high neutron flux trip for the ' " -* *"y feedback cases. The neutron flux level in the-core rises rapidly for these insertion rates while core heat flux and coolant system temperature lag Dehind tiue to the thermal capacity of the fuel and coolant. system fluid. Thus, the reactor is trippeo prior to significant increase in heat flux or water temperature with resultant high minimum DNB ratios during the transient. As reactivity insertion rate decreases, core heat flux and coolant temperatures can remain more nearly in equilibrius with the neutron flux; minimum ONS ratio during the-transient thus decreases with decreasing insertion reten fo* fhe re moye en dech A e. hs3h Alex 4r'p pro vide; f r me ry fire feefien .
- 2. The overtemperature ai reactor trip circuit initiates a reactor trip when-measured coolant loop AT exceeds a setpoint based on measured Reactor Coolant System average temperature and pressure. This trip circuit is described in detail in Section 7.2.1. It is impo and average temperature contributioru to the circuit %(tant to compensated 1ead-lag note that the leer 4T in order to decrease the effect of the thermal capacity of the Reactor Coolant System in response to power increases.
assuming perssuro a.er fuessure c o n &* / of t r" b l e->
- 3. With a further decrease in reactivity insertion rate,'the overtemperature AT and high neutron flux trips become equally ef fective in terminating the p transient (e.v., n .;m ;; ' w m , m q v i Ly inm i s,. rn:). For Mest p reactivity insertion. rates astr=r - E - 10 5 AW e W ' ' "
- IM L' u the effectiveness of the overtemperature AT trip increases (in terms of 6' increased minimum DNB ratio) due to the fact that with lower insertion rates the power increase rate is slower, the rate of increase of average
( coolant temperature is slower and the system lags and delays become less significant. {s 3, c hve.s aut pqe)
~
D e-ca qtivity insertion rates less than s 1.5 x 10 5 Ak/sec, the 4se "in the reacttr-coolant temperature is sufficiently hi Ltha e steam f generator safetyNve-.4G' point is reacheJL.pekt o trip. Opening of these valves, which act as anM innal' heat sink for the Reactor Cool-ant System, snarply oecreases-t'h'e r crease of Reactor Coolant System average temperat'u're. This decrease e-of increase of the average coojant'iystem temoerature during the transienh-acqntuated by l the leae*1ag compensation causing the overtemperature aT trip set point to b
/ e4eeched later with resulting lower minimum DNB ratios. \
m . ' ( replace. wilh insert frem a4 ached fog e l 15.4-8
'005
. ";t:*0
- _ - - . . . - . _ - . - . - - - _ - - - - - - _ _ - - ~ _ . - - - - - . - -
s t With a further decrease in reactivity insertion rates, assuming pressurizer pressure control inoperable, the high neutron flux and high pressurizer pressure trips become equally effective in terminating the transient. The higher pressures reached due to the assumed inoperability of normal pressure control tend to increase the DNBR. However, these higher pressures reduce the
- effectiveness of the overtemperature AT trip discussed in the l previous paragraph. This is because the overtemperature AT i' setpoint equation has a term which increases when measured pressurizer pressure exceeds the reference pressurizer pressure. j A sensitivity study was therefore performed to ensure that the !
minimum DNBR ls abovi the limit value for ecch pressure control availability assumption.
- 4. For even slower reactivity insertion rates, the steam line safety ;
valves might lift prior to reactor trip. Whether this occurs < depends on several factors, including the steam line pressure at the beginning of the transient, the safety valve actpoints, the amount and type of reactivity feedback, and the availability of the steam line PORVs. The effect of steam line safety valve and i PORV-lift depends upon the reactivity insertion rate and feedback. In. general, valve lift tenda to reduce the rate of increase of reactor vessel average temperature, which is a component of the overtemperature AT trip setpoint equation. For this transient, the lead / lag compensation in this equation causes this increasing temperature to decrease the setpoint, making a trip more likely. Even if no credit is taken for this compensa-tion, a conservative assumption which bounds the effect of valve lift, the maneuvering analysis described in Reference-12 onsures that the minimum DNBR remains above the limit value for insertion rates slow enough to trip on overtemperature AT. r 1 i l: l
~
l-l
f ttJures n.e.2-7,15 t.-2-8 ana1ST472r9 llloitrateminimum-BM11LcalculataMer minimum-a ne -we n eum-re scrt vi tyM e eco a c r. Since the RCCA withdrawal at power incident is an overpower transient, the fuel temperatures rises during the transient untti af ter reactor trip occurs. For j high reactivity insertion rates, the overpower transient is f ast with respect ' to the fuel rod thermal time constant, and the core haat flux lags behind the neutronfgv,, response. Due to this lag, the peak core heat flux does not ; exceed 4M percent of its nominal value (i.e. , the high neutron flux trip ' setpoint assumed in the analysis). Taking into account the effect of the RCCA i withdrawal on the axial core power distribution, the peak fuel temperature will ' still remain below the fuel melting temperature. ; For slow reactivity insertion rates, the core heat flux remains.more nearly in equilibrium with the neutron flux. The overpower transient is;terininated by the overtemperature AT reactor trip befor a ONS condition is reached. The-Deak' heat flux again is maintained below percent of its nominal value. l Taking into account the ef fect of the RCCA withdrawal on the axial core power- l distribution, _the peak centerline temperature will remain below the fuel melting temperature. Since ON8 does not occur at any time during the RCCA withdrawal at power tran-sient, the aDility of the primary coolant to remove heat from the. fuel red is not reduced. Thus, the fuel cladding temperature does not rise.significantly , above its initial value during the transient. The calculated sequence of events for this accident is shown on Table 15.4.1-1; With the reactor tripped, the plant eventually returns to a stable condition. The plant may subsequently be cooled down further by following normal plant shutdown procedures. 15.4.2.3 Environmental consecuences . The reactor trip causes a turbine trip, and heat is removed from the seconaary system througn the steam generator power relief valves or safety valves. Since l no fuel damage is postulated to occur, the raciological consecuences associated with atmospheric steam release from this event would be less severe than the
- steamline break accident analyzed in Section 15.1.5.
15.4.2.4 Conclusions hi The hign neutron flux,^gh pres suriece pressure, ind overtemperature AT trio channels provide adeauate protection over the entire _ range of possible reactivity insertion rates, i . e. , i the minimum value of DNBR is always larger than the limit value. The radio-I logical consecuences would be less severe than the steamline break accident analyzed in Section 15.1.5. L s t 15.4-9 190fr+,Huta
15.4.3 R00 CLUSTER CONTROL ASSEMBLY MISOPERATION (System Malfunction or Operator Error) 15.4.3.1 Identification of Causes and Accident Description : Rod cluster control assembly (RCCA) misoperation reaidents include:
- 1. One or more dropped RCCAs within the same gro w
- 2. A dropped RCCA bank
- 3. Statically misaligned RCCA
- 4. Withdrawal of a single RCCA .
Each RCCA has a position indicator channel which displays the position of the assemoly. The displays of assembly positions are grouped for the operator's convenience. Fully inserted assemblies are further indicated by a rod at-bottom signal, which actuates a local alarm and a control room annunciator. Group demand position is also indicated. Full length RCCA's are always moved in preselected banks, and the banis are-always moved in the same preselected sequence. Each bank of RCCAs is divided into two groups. The rods comprising a group operate in parallel through multiplexing thyristors. The two groups in a bank move sequentially such that the first group is always within one step of the second group in the bank. A definite schedule of actuation (or deactuation of the stationary gripper, movable gripper, and lift coils of a mechanism) is required to withdraw the RCCA attacned to the mechanism. Since the stettonary gripper, movable gripper, and lift coils associated with the four RCCA's of a rod group are driven in parallel, any single failure which would cause rod withdrawal-would affect a minimum of one group. Mechanical failures are in the direction of insertion, or immobility. The cropped RCCA assemblies, dropped RCCA assembly bank, and statically mis-aligned RCCA assembly events are classified as ANS Condition II incidents (incidents of moderate frequency) as defined in Section 15.0.1. The single RCCA withdrawal incident is classified as an A#- Condition III event #8 diseussed below. (pgem[3{.for raam No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single RCCA from the inserted bank at full power operation. The operator could withdraw a single RCCA in the control bank since this feature is necessary in' order to retrieve an assembly should one be accidentally dropped. The event analyzed must result from multiple wiring failures-(probability for single rancom f ailure is on the order of 10 s/ hour-refer to Section 7.7.2.2) or multiple significant operator errors and subsequent and repeated operator disregard of event indication. The probability of such a combination of conditions is considered low enough that the limiting consequences may include slight fuel damage. Thus, consistent with the philosophy and format of ANSI N18.2, the event is classified as a Condition III event. By definition "Cendition III occurrences 15.4-10 1^CC Uv e be
! i i urgent f ailure alarm also inhibits automatic rod motion in the group in which it occurs. Withdrawal of a single RCCA by operator action, whether deliberate or by a combination of errors, ould result in activation of the same alarm and the same visual indications. Withdrawal of a single RCCA results in both a positive reactivity insertion tending to increase cos e power and an increase in local power density in the core area associated with the RCCA. Automatic protection for this event is provided by the overtemperature AT reactor trip, j although, due to the increase in local power density, it is not possible in all cases to provide assurance that the core safety limits will not be violated.. 15.4.3.2 Analvs1s, of Effects and Consecuences
- 1. Oropped RCCAs, dropped RCCA bank, and statically misaligned RCCA.
Method of Analysis p.
- a. One or more dropped RCCAs from the same group.
For evaluation of the dropped RCCA event, the transient system response is calculated using the LOFTRAN code. The code simulates. the neutron kinetics Reactor Coolant System, pressurizar, pressurizer relief and safety valves, pressurizer spray, steam ! generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level. _ Statepoints are calculated and nu'elear medels are used to obtain as hot channel, factor consistent with the primary system conditions anC reactor power. By incorporating the primary conoitions from the transient and the hot channel factor from the nuclear analysis, the-DN8 design basis is shown to be met using the THINC code. The transient response, nuclear peaking factor analysis, and DNB design basis confirmation are performed in accordance with the methodology _ described in Reference 6.
- b. Statically Misaligned RCCA Steady state power distributionsare analyzed using the computer codes as described in Table 4.1-2. The peaking f actors are then wee-** c.ogore.d
!7"* +a the ,IMNC~ cede ,;p ;;;;;;;t; M: C. :
to new nd p.Kig t,a 6 fr.gm Results VIPRE -oI N lRhc.e Q*
- a. One or more Oropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion which may be detected by the power range negative neutron flux rate trip circuitry. If detected, the reactor is tripped within approximately 2.5 seconds following the drop of the RCCAs. The core is not adversely affected during this period, since power is decreasing rapidly 4 Following reactor trip, normal shutdown procedures are followed. The operator may manually retrieve the RCCA by following approved operating proceoures.
1 i l 15.4-12 E00 W e;;
mhdm3 RC4A posit >on unceriad*8f for this RCCA m salignment, with bank 0 inserted to its full power insertion limit and g one RCCA fully withdrawn, DNBR coes not fall celow the limit salue, this case is analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values M-i; c;... r n; (as given in Table 15.0.3-4)).we with the increased radial peaking factor associated with the miseligned g ggL igd kMas t esc *Ad for Mae SCO e4ab Lmet DNB celculations have not been performed specifically for RCCAs missing from other eanks. However, power shape calculations have been done as required for the RCCA ejection analysis. Inspection of the power shapes shows that the DNB and peak kw/ft situation is less severe than the bank D case discussed above, assuming insertion limits on the other banks equivalent to a bank 0 full-in insertion limit. - For RCCA misalignment $ with one RCCA fully inserted, the ONBR does not fall below the limit value. This case is analyzed assuming the-initial reactor power, pressure, and RCS temperatures are at their nominal values, including uncertainties (as given in Table 15.0.3-4) but with the increased racial peaking factor associated with the misaligned RCCA. DNB does not occur for the RCCA misalignment incident and thus the-ability of the primary coolant to remove heat from the fuel red is not reduced. The peak fuel temoerature corrdsponds to a linear heat generation rate based on the racial peaking factor penalty associated with the misaligned RCCA and the design axial power distribution. The resulting linear heat generation is well below that which would , cause fuel melting. Following the identification of a RCCA group misalignment condition by the operator, the operator is required to take action as required by the plant Technical Specifications and operating procecures.
- 2. Single RCCA Withdrawal Method of Analysis Etowe.e,. w h miu t frei
- ohN 088ta,5 Power distributions within the core are calculated using the computer ccdes as described in Table 4.1-2. The peaking factors are then used by THINC to calculate the DNBR for the event. The case of the worst rod
withdrawn from bank 0 inserted at the insertion limit, with the reactor initially at full power, was analyzed, This incident is assumed to occur at beginning-of-life since this results in the minimum value of moderator temperature coefficient. This assumption maximizes the power rise and minimize the tencency of increased moderator temperature to flatten the power distribution.
~
I 15.4-14 O iv h 4
The single RCCA withdrawal transient is analyzed by employing the , RETRAN-02_ computer code (Reference 1). The code simulates i neutron kinetics, decay heat, the Reactor Coolant System (RCS), [ control rods, pressuriter, pressurizer power-operated relief l valves (PORVs), pressurizer spray, steam generator, turbine, and the Reactor Protection System. The code computes pertinent plant ; variables including power level, temperatures, pressures, mano i flow rates, and liquid inventories. The DNBR calculation for this accident is performed with the VIPRE-01 computer code (Reference 6) using the Statistical Core Design pr'ocedure described in Reference 8. The result of this calculation is a power peaking limit which, at the statepoint, yields a DNDR equal to the limit value. All fuel pins exceeding this peaking limit are assumed to undergo DNB and subsequently ! fail, t Plant characteristics and initial conditions are discussed in Section 15.0.3. In order to obtain conservative results for a : single RCCA withdrawal accident, the following aJaumptions are - made: -
- 1. The values assumed for initial reactor power, pressurizer i pressure, RCS average temperature, and RCS flow include no allowance for uncertainties. Uncertainties in initial conditions are included in the SCD DNBR limit as described in Reference B.
- 2. A most negative moderator density coefficient-of reactivity is assumed corresponding to the beginning of core life. A least negative variable fuel temperature coefficient is assumed corresponding to thn beginning of core life. These f eedback assumptions lead to the most limiting statepoint with respect to trermal-hydrau. tic conditions. Peaking -
limits derived from this:statopoint are compared to peaking , results from beginning and end of core life to ensure that ! the percent of fuel rods in DNB is smaller than the ; acceptance criterion. L
- 3. The reactor trip on high neutron flux is assumed to be 'I actuated at a conservative value higher than 113.2 percent '
of nominal iull power. The calculated neutron flux is conservatively modified to account-for transient changes in l the-flux incident on the excore detectors. The AT trips l include all adverse instrumentation and-setpoint errors. The. delays for trip actuation are assumed to be maximum values.
- 4. The RCCA trip insertion characteristic is based on the '
assumption *that the highest worth assembly is stuck in its fully withdrawn position.
- 5. The maximum positive insettion rate is greater than that for ..
the maximum speed withdrawal of the most reactive single Control Bank D RCCA from at or above its insertion limit, accounting for uncertainties in the indicated RCCA position. - 1.
- 6. The case presented assumes normal pressurizer spray operable but assumes the pressurizer PORVs inoperable. A sensitivity
--,,--,-mme nn-r-n,m-rn m m.w.. -
--.e,,-.,,w,,,m.,,--n.-,c,,rw.wm-n..m.,m emp ,m yy,.,.n- .,myn,wm ,,swp,-..e ---w,-n,,qu-gm yn
. _ . _ _ _ _ _ . m______-. . . _ - ..- _m . _ - . - . _ . . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _
l 1 1 l l i study was perform (<d to ensure that the minimum DNDR for this case bounds other pressurizer pressure control availability . assumptions. E I i V h d
?
w f e 6 T t s b 9 9 t-I . p
Results b5cd ONhed I rod withdrawal event, two cases Mye been considereo as For the single follows:
- a. If the reactor is in the manual control moce, continuous withdrawal of a single RCCA results in both an increase in core power ano cool-ant temperature, and an increase in the local hot channel f actor in the area of the withdrawing RCCA. In terms of the overall system ,
response, this case is similar to those presented in Section 15.4.2. ; However, the increased local power peaking in the area of the with-drawn RCCA'results in lower minimum DNBR's than for the withoraw) bank cases. Depending on initial bank insertion and location of the withdrawn RCCA, automatic reactor trip :ay not occur f ast enough to prevent the minimum core ONBR from fa ing below the limit value; Wever' - on OTWT or mea pru miser prua rg. M '1:ti;r, ;f ;,i; : n : ;t tM ;:urr ud :=ir t : -d4tiem it *
^^ ~;;r;;ge, e m .:. I t " - Mdb ::'; :ted t: tH, the pl = t 2:cs t.ht an upper limit for the number of rods with a ONBR less than the limit value is 5 percent.
- b. If the reactor is in the automatic control mode, the multiple failures that result in the withdrawl of a single RCCA will result in the immobility of the other RCCAs in the controlling bank. The transient will then proceed in the same manner as Case (a.) described above. '
For such cases a reactor trip wi ultimately occur as above, although not fast enougn in ali cases to prevent a minimum DNBR in the core of less than the limit value. Following ieactor trip, normal shutdown procedures are followed. 15.4.3.3 Environmental Consecuences M The most limiting rod cluster control assemoly misooeration, accidental .ith-d g drawal of a single RCCA, is predicted to result in less than IP. fuel clad
- M **T damage. The subsecuent reactor and turbine trip would result in atmospheric s
- 2. steam cump, assuming the condenser was not available for use. The radiolog-ical consequences from this event would be no greatcc than the main steamline break event, analyzed in Section 15.1.5. 1 L
i 15.4.3.4 -Conclusions For cases of dropped RCCAs or cropped banks, for which the reactor is tripped , i by the oower range negative neutron flux rate trip, there is no reduction in the margin to core thermal limits, and consequently the ONB design basis is met. It is shown for all cases which do not result in reactor trip that the DNBR remains greater than tne limit value and, therefore, the ONB design is met. ! For all cases of any RCCA fully inserted, or bank 0 inserted to. its rod u insertion limits with any single RCCA in that bank fully withdrawn (static misalignment), the DNBR remains greater than the limit value. 15.4-15 5 6 :=
Attachment 1: ! Table 15.4.1-1 shows the sequence of events for the single
- uncontrolled rod withdrawal transient. Figures 15.4.3-3 through 15.4.3-6 show the transient response for the single uncontrolled rod '
withdrawal event. System temperature and system pressure increase until reactor trip occurs, which is after the RCCA is completely t withdrawn, i f 1 i i i-I N i 4 h I i h o f
)
i i 2 0 - E i t E 9
)
1 s
= w w ,, o m. =*- +- ,e e> --,-u + e + -
w+ <w+-w- y= sews- e im m +- ..r , --ww. r w- e nag 4 ar vryg er e- e- .-+. w ypue- ,g w ,p y. g - - t er=-c '.e-
l l l Attachment 2: 15.4.3.3 Environmental consequences The postulated rod cluster control assembly misoperation (with the litniting case being the uncontrolled withdrawal of a single RCCA) causes a reactor and turbine trip, which subsequently results in atmospheric steam dump. For conservatism, the condensate system is assumed to be inoperable, thereby limiting the amount of Jodine removal prior to atmospheric steam dump. The atmospheric steam dump does not result in the release of any radioactivity unless there is leakage from.the RCS to the secondary system in the steam generators. A conservative analysis of the potential offsite doses resulting from this accident is presented considering equilibrium operation based upon one percent defective fuel and a 1.0 gpm steam generator leak rate prior to the accident. At the time of the accident, 5 percent of the fuel is assumed to fall, thus releasing a significant amount of the gap inventory to the RCS. Based on this, the following two cases are analyzed Case 1 Normal equilibrium Technical Specification iodine , concentrations exist at the time of the accident (see Tabic ' 15.0.9.2). Case 2: There is a pre-existing iodine spike at the time the accident occurs. The reactor coolant, concentrations are the maximum permitted for full power operation; i.e., 60 times the normal concentrations. The following assumptions and parameters are used to calculate the , activity release and subsequent offsite doses for the postulated RCCA misoperation:
- 1. The total primary-to-secondary leak rate (PSL) is 1.0 gpm. For '
the duration of the accident, this leak rate is evenly
-distributed between the four steam generatorst i.e., 0.25 gpm PSL in each steam generator.
r
- 2. Offsite power is lost when the transient occurs. .
- 3. All noble gases which leak to the secondary side are released via >
the steam release without any reduction due to removal mechanisms. 4 The iodine partition factor during the accident is 0.01.
- 5. Other assumptions are listed in Table 15.4.3-1.
Based on the model described in Appendix ISA, the thyroid and whole body doses are calculated at the exclusion area boundary and the low population = tone. . The results are-presented in Table 15.0.7-2. The doses are within the limits of 10 CFR-100. k P
.,--,,,.w, --,,-,,-,,-,-..vv...,,,..,,,,,,-c p,_ ,-,.%, .w,,v.,y...
1 1 In order to dilute, two separate operations are required. ;
- 1. The coe'ator must switch frorn the automatic makevo mooe to the dilute mode
- 2. The start cutton must be depressed.
Omitting either steo aoulo prevent ailution. Information on the status of the reactor coolant makeup is continuously available to the coerator. Lights are provided on the control board to indicate the operating condition of the pumps in the Chemical and Volume Control System. Alarms are actuated to warn the operator if boric acid or demineralized water flow rates deviate from preset values as a result of system malfunction. The signals initiating these alarms also cause the closure of control valves terminating the addition to the Reactor Coolant System. A boron dilution is classified as an ANS Condition II event, a fault of moderate frecuency. See Section 15.0.1 for a discussion of Condition 11 everts. 15.4.6.2 Analysis of Effects and Consecuences Method of Analysis Boron dilution during refueling, startuo, and power operation is considered in this analysis. jggeyg hun 6 gg I~ Dilution during Refueling n
'N efueling the foliowing conditions normally exist: ,e Durinpr\
- 1. Oneresidualheatremovalpumoisoperatingtoensurecont.iNous mixing irNhe reactor vessel. /
\ !
- 2. The reactor makaup control system is adjusted for ad6ition of boric acid solution at'f'efueling concentration.
N
- 3. The charging pumps are inoperative. ,/'
y ,
- 4. The baron concentration in the refueling water is approximately 2000 ppm, corresponding to a shutdow Amargin of at least 5 percent ak/k with all rod cluster control asswiblies out. Periodic samoling l ensures that this concentration'is '
maintained. I
- 5. Since neutron sources are' installed ', theN ore, the source range detectors outside thefreactor vessel are acti've and provide an audible count rate /During initial core loading 1F3 detec+ ors are installed inside/t he reactor vessel and are connecteq to instrumentation'giving audible count rates to orovideNirect monitoring,91' the core. N,
/
- 6. A higryflow alarm at the discharge of the CVCS (f rom flow element NVFy S530) is active providingaan alarm to the operator wnen the f ow f te f from the charging pumps exceeds 175 gpm.
/ ;
a- ~~ 15.4-19 _,m
.-~ . - . - - _ - - - - _ - - _. _. - - . - - . . - - - - - . _ - - . _ - - . -
Alarm Function Which Initiates Mittaation Mitigation of a boron dilution accident is not assumed to oegin until an alarm has warned of the abnormal circumstances caused by the event. For Mode 6, the alarm function is provided by the source range high-flux-at-shutdown alarm exceeding its setpoint. For Mode 2 and for , manual rod cantrol during Mode 3, the alarm function is provided by ! I the earliect reactor trip setpcint reached. Finally, for automatic rod control during Mode 1, the alarm function is provided by the alarm ' l which occurs when the' control rods reach their insertion limits. Dilution Voluma A postulated dilution event progresses faster for smaller RCS water volumes. Therefore, the analysis considers the smallest RCS water volume-in'which the unborated water is actively mixed by forced circulation. for Modes 1 and 2, the Technical Specifications require - that at least cne' reactor coolant pump be operating. This forced circulation wi14 mix the RCS inventory in the reactor vessel and each of the four reactor coolant-loops. The pressurizer and the pressurizer surge line are not included in the volume available for dilution in Modes 1 and 2. For Mode 6,-the reactor coolant water level'may be drained to below the top of the main coolant loop piping, and at least or.e train of the Residual Heat Removal System (RHRS) is operating. The volume available for dilution in this mode is limited to the smaller valume RHRS train plus the portions of the reactor ' vessel and reactor coolant loop. piping below the minimum water level (8 inches above the centerline of the hot and cold leg piping) and between the RHRS inlet and outlet connections. The minimum water level used to calculate this volume is corrected for level instrument uncertainty. Boron Concentrations The Technical Specifications require that the shutdown margin in the various modes be above a certain minimum value. The difference in boron concentration, between the value at which the relevant alarm 1 function is actuated and the value at which the reactor is just critical, determines the time available to mitigate a dilution event. Mathematically, this tine _ is a function of the ratio of these two concentrations, where a-large ratio corresponds to a longer time. During the reload safety analysis for each new core, the above
~
concentrations are checked to ensure that the value of this ratio for i each mode is larger than the~ corresponding ratio assumed in the-accident analysis. Each mode of operation covers a range of
-temperatures. Therefore, within that mode, the temperature which-minimizes this ratio is used for comparison with the accident analysis ratio. For accident initial conditions in_which the control rods are
" withdrawn, it is conservatively assumed, in calculating the critical
~ boron concentration,=that-the most reactive-RCCA-does not-fall into-
- the-coreJat reactor trip. -
1 l7-l Dilution Flow Mate In the absence of flow rate restrictions, the dilution flow rate assumed to enter the RCS is greater than or equal to the desion s-r*- . . .e- .-.w .-1.------wn--*-- -- erm -,we e,m---w-.r e wwei e,,, ,-r ,ie..w ,4% .,,e,,,m.z-.e---,.-.mme -, 3 w e v.r ,wp., , y- w-gy,
h j volumetric flow rate of both reactor nakeup water pumps. In a
' dilution event, these pumps are assumed to deliver unborated water to the suction of the centrifugal charging pumps. Since the water delivered by these pumps is typically colder than the RCS inventory, the unborated water expands within the RCS, causing a given volumetric i flow rate nwasured at the colder temperature to correspond to a larger volumetric dilution flow rate within the RCS. This density difference in the dilution flow rate is accounted for in the analysis. The above :
assumption on flow rate is also conservatively used for Mode 6, even though the dilution flowpath in the Chemical and Volume Control System i (CVCS) is isolated during refueling (normally by locking valve NV-250 closed). This isolation prevents unborated makeup from reaching the RCS. Any makeup which is required during this mode is borated water supplied from the refueling water storage tank. Results The calculated sequence of events is shown in Table 15.4.1-1. 4
- Dilution Durina Refuelino (Mode 6)
During Mode 6 an inadvertent _ dilution from-the Reactor Makeup Water _ System is prevented by administrative controls which isolate the RCS from potential sources of unborated makeup water.- _The results _ presented in Table 15.4.1-1 for this mode are for an-assumed dilution event, for which no mechanism or flowpath has been identified. .The results presented in Table 15.4.1-1 show that,-with a limitation on the flow rate from potential sourdes of unborated water, . there is adequate time for the operator to determine the cause of the dilution,
. isolate the_ source of unborated water, and initiate reboration before the shutdown margin is exhausted. Adequate !ime is judged to be at least 30 minutes for Mode 6. The results presented in Table 15.4.1-1 are-for.the dilution flow rates which, assuming the boron concentration ratios are at the reload safety analysis _ limits, give exactly these operator response times. Flow rates _are restricted, through Technical Specifications and administrative controls, to values which are less than these analyzed flow rates, thus in practice giving even longer operator response times. Additional margin is provided by the fact there is typically margin between the_ assumed boron concentration. ratio for a given mode and the actual corresponding concentration ratio ~for the reload core, r
9
\ ,,c- ,n n n . < - - , , - - , - , ,---+- =,,-,w- ,---u. ,-~w w e~ , . ,=m-.
_ .= -_ _ - - _ _ . -
. - -. __. ~_._ . . __ __ . _. .-. . _ _ - %\de. m s
A minitrum aster volume in the Reactor Coolant system is considered. This / cogresponcs to the volume necessary to fill the reactor '.essel 3 cove the / noI2jes to erisure mixing via the residual heat removal loop. / The a\nalysis assumes a 200 gpm acdition of uncorated water f rem the Referring to Figures 5.3.4-2 ano Chemicaf\sno volume Control System. 9.3.4-3 (1 of 4), three cotential paths exist f or introauction of react'or makeup adtet to the suction of the Charging Pumps. The first is the,4wo incn line inbluding valve NV-262, nich is normally closed. '
,/
/
The other two paths are througn 2 inch lines, one of which leaas' to the volume control tank with the other bypassing this tank. Thes e' l i nes contain flow contrb.) valves NV-171A and NV-175A respectively'
\
Inherent in the assumption of 200 gpm delivery of unborated water through these paths are the following necessary conditions: '
- 1. Theoperationofano'rmallyinoperativechargiNpumo,
/
- 2. The operation of the normally inoperative r/ actor 1akeuD water subsystem, at sufficient capacity to provide 200 gpm to the CVCS,
/
- 3. An open path from the reactors makeup water subsystem to the charging pumps suction, j/
Thus, in order for dilution to occur,'4n operator woula nave to take several independent actions conscious 19\ directed to water makeup. He fs would have to set the mode selector / witch to " Manual" or ")ilute", the
) ers to the desired flow rates, the boricacidandmakeupflowcontro)integratorstothedesiredQuantities, boric acid and makeup water batctk actuate the start switch, and f Mally start as charging pump. The flow alarm discussed in Item 6 orov) des an additional indication to the operator if the dilution floV rate into the Reactor Coolant f.jstem exceeds the values assumed in this, analysis. '
e The above sequence of events can not occur independently. The only way l they can occur is through conscious action by the operator. In addition, l for the next 57 minutes, the operator and refueling personnel mest be l oblivious to any incications of a dilution, including a continually rising audible count rate / from the source range NIS,
/
In addition to the unit alarms and core monitoring systems used to provide protection agafnst boren cilution accidents during initial core loaoing or - i refueling, aosinistrative controls .nicn isolate the Reactor Coolant i System from/the potential source of unborated water are also employed. l_ Valve NV-250 is locked closed during refueling operations. This valve blocks the potential flow paths **hicn could allow unborated makeup water to reactI the Reactor Ccolant System. Any makeup which is reauired'during i refueling is supplied f rom the refueling water storage tank. Thus , 'it is ! concfuded that appropriate controls have been included in the design to
! mirtimize the probability of this event occurring and to allow adeouates l
tise for operator action (57 mintd.es) in the unlikely event that dilution ' hould occur during refueling. 15.4-20 ,. . .ae
kWh. _
'lilution During Startup
/
Core N monitoring is oy external SF3 detectors. Mixing cf the reactor / ~ coolantxis accomplisned by operation of the reactor coolant pumps. High source range flux level and all reactor trip alarms are effective.
'N / .
In the analysh, a conservatively high dilution flow capacit for the two-primary water makeup pumos is considered. The volume of the reactor core. is the active voluk of the Reactor Coolant System excMding the . prossurizer. / %, Oilution at Power ,e y di T y
'N~ / '
D With the unit at power and the' Reactor Coorant System at pressure, the dilution rate is limited by the. capacity'of the charging valve (analysis:. is performed assuming all charging humps are in operation, although only -- one is normally in operation). The' eMective reactivity addition rate 1say,' a function of the reactor coolant temperat.ure and boron concentration. Tf The reactivity insertion rate calculated is b'ased on conservatively high-value for the expected #6ron concentration at p'ower as well as a conservar tively high charging flow rate capacity. Results . g Dilution Our'ino Refueling [ Oilut.itn during refueling cannot occur without the conscious action of the. oper'a to r. Several independent actions are required by the operator s for e
,tfie dilution to occur. However, for purposes of the analysis, these N Even so, there is aoeauate time (greate' e'/ actions are assumed to occur.than 57 minutes) for the operator to recognizs the high count r
[ and terminate the dilution. Dilution During Startuo et M t For dilution during startuo there is adequate time (_7 n :- ".18' , minutes) for the operator to recognize the high count rate signal and manually terminate the source of dilution flow, l Oilution at Power d Itast 65 wweseos
,/ *
- 1. With the reactor in automatic control, the' power and temperature increase from coron dilution results in,Ansertion of the rod cluster-control assemolies and a decrease in the shutdown margin. The rod insertion limit alarms (low and low-Tow settings) provide the operator with adeauate time (J _ , 1 ~ , X ~ _ ::) to cetermine the cause of dilution, isolate the primary grade water source, and l
initiate reboration before the total shutdown margin is lost due to l dilution. l l 15.4-21 L 3930 We _ __ _
1
- 2. With tne reactor in manual control and if no operator action is taken, the power and temperature rise would cause the reactor to ,
reach the overtemperature AT trip setpoint. ^ ^ ^ ' a" l
""'1., ..:, -
l 37 t ,,, m i
.t. ;-
7' .'w, nn "_ : : ' _' - ;;-i - " " -; 2 J um--- y * *i;- - ,(j,l ;- m4- tm --; , ,, _ _rj : --+n u mithin m . - s v. *1-- __,,,,mm, Prior to the overtemperature AT trip an overtemperature ai alarm and be actuated. There is adequate time available l turbine runback -z would
.9 ^^^) after a recctor trip for the operator to.
determine the cause of dilution, isolate the primary grade water sources and initiate reboration before the reactor can return to , j6 gnie8f88 15.4.6.3 Environmental Consequences ; : 1 There would be minimal radiological consequences associated with a Chemical and-Volume Control System malfcnction that results in a decrease in boron con- 4 centration in the reactor coolant. The reactor trip causes a turbine trip, and' heat may be removed from the secondary system through the steam generator power-relief valves or safety valves. Since no fuel damage occurs from this transient, the radiological consequencer, associated with this event are less . severe than the steam line break event enalyzed in Section 15.1.5. s 15.4.6.4 Conclusions b Although the above results do not explicitly demonstrate fulfillment of the Standard Review Plan requirements for all modes of plant operation permitted by the Technical-Specifications, Reference 7 states that these requirements ! will not be backfitted to operating plants. The results presented above show that there is adequate time for the operator to manually terminate the source, of dilution flow. Following termination of the dilution flow, the reactor will be in a stable condition. The operator can then initiate reboration to recover-the shutdown margin. , l 15.4.7 IhADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSIT 0N 15.4.7.1 Identification of Causes and Accident Description I Fuel and core loading errors such as can arise from the inadvertent loading of one or more fuel assemblies into improcer positions, loading a fuel rod during' manufacture with one or more pellets of the wrong enrichment or the loading of a full fuel assemoly during manufacture with pellets of the wrong enrichment will lead to increased neat fluxes if the error results in placing fuel in core positions calling for fuel of lesser enrichment. Also included among possible core loading errors is the inadvertent loading of one or more fuel assemblies requiring burnaDie poison rods into a new core without burnable poison rods. l Any error in enrichment, beyond the normal manufacturing tolerances, can cause power shapes which are more peaked than those calculated with T.he correct enrichments. There is a 5 percent uncertainty margin included in the design value of power peaking factor assumed in the analysis of Condition I and 15.4-22 2 " i W "-
. - - _~ . - _ _ _ . - _ . -
l Condition 11 transients. The incore system of moveaole flux cetectors, .hich is used to verify Dewer shapes at the Deginning of cycle, is capable of reveal) g . any absembly enrichment error or loading error whien causes r.ower snace!. to ce l peaked in excess at the design value. To reduce the probaD11ity of core loading errors, eacn fuel assemoly is marxed with an identification num0er ana loaded in accoraance with a core loading di a-gram. During core ioading, the identification aumoer will be checked bef ore each assembly is moved into the core. Serial numbers read Juring fuel movement we subsecuently recorded on the loading diagram as a further check on proper l placing after the loading is completed. The power distortion due to any combination of misplaced fuel assemblies would significantly raise peaking f actors and would be readily observable with incore flux monitors. In addition to the flux monitors, thermocouples are located at the outlet of about one third of the fuel assemblies in the core. There is a high probability that these thcrmocouples would also indicate any abnormally high coolant enthalpy rise. Incore flux measurements y ' e" during the startup subsecuent to every ref ueling operation. and Cors f.pmatry chee.ks nee, Pe rfoiws.d The inadvertent 11ading and operation of a fue! assembly in an improper position is classified at an ANS Condition III event, a infrequent f ault. See Section 15.0.1 for a discussion of Condition III events. 15.4.7.2 Analysis of Effects and Consecuences Method of Analysis Steady state power distribution in the x y plane of the core ara calculated using the computer codes as described in Table 4.1-2. A discrete representa-tion is used wherein each individual fuel rod is described by a mesh interval. The power distributions in the x y plan for a correctly loaaed core assemoly are also given in Chapter 4 based on enrichments given in that section. For each core loading error case analyzed, the percent deviations f rom aecector readings for a normally loaded core are snown at all incore detector locations (see Figures 15.4. 7-1 to 15.4. 7-5, inclusive). Results The following core leading error cases have been analytea: Case A: Case in which a Region 1 assemoly is interchanged with a Region 3 assemol? The particular case considered was the interchange to two adjacent asse.*;ies near the per4pnery of the core ( see Figure 15.4. 7-1). L l Case B: l Case in hich w a Region 1 assembly is interchanged with a neighboring Region fuel assemoly. Two analyses have been performed for this case (see figures 15.4.7-2 and 15.4.7-3). l 15.4-23 '-~~"N*
-~ . -- -. _ _ _ - - . _ -- - . - __- -_ = - .
15.4.8.3 Environmental Consecuences The following conservative assumptions are used in the analysis of t'1e release of radioactivity to the environment in the event of a postulated rod ejection accident. A summary of parameters used in the analysis is given in table 15.4.8-2.
- 1. Ten percent of the gap activity is released to the containment atmosphere.
- 2. 5 percent of the gap icaine activity is released to the containment sump and available for release via ECCS leakage outside containment.
- 3. Annulus activity which is exhausted prior to the time at which the annulus reaches a negative pressure of -0.25 ir..w.g. is unfiltered.
4 ECCS leakage occurs at twice the maximum operational leakage.
- 5. ECCS leakage begins at the earliest possible time sump recirculation can begin.
- 6. Bypass leakage is 7 percent.
- 7. The effective annulus volume is 50 percent of the actual volume. ,
- 8. The annulus filters become fouled at 900 seconds resulting in a 15 percent reduction in flow.
- 9. Elemental iodine removal by the ice condenser begins at 600 seconds and continues for J.H0 seconds with a removal efficiency of 30 percent.
1569
- 10. One of the containment air return fans is assumed to fail,
- 11. The containment leak rate is fif ty percent of the Technical Specifications limit after 1 day, at
- 12. Iodine partition factor for ECCS leakage is 0.7 for the course of the accident.
- 13. No credit is taken for the auxiliary building filters for ECCS leakage.
t The redundant hydrogen recemoiners and igniters fail. Therefore, purges l 14. are required for hydrogen control.
- 15. Equilibrium is reached after 200,000 seconds such that the only annulus ventilation discharge is due to inleakage.
- 16. Water density at 160'F is used to calculate the sump water mass following the accident; l 17. Elemental and particulate iodine removal from the containment atmosonere by containment spray begins at tnc iniation of the event and continues y for 112 minutes yith spray removal LA.BDAS "
of 1.52.E-02 minutes 1 and ( 4.10E-02 minutes t, respectively. I 15.4-33 a:lCC
_ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ - . _ _ _ ~ _ _ _ _ _ . - _ . __ _. __.. ... _ _ _ _._ REFERENCES FOR SECTION 15.4 .
- 1. 10 k= , D_ u ,-j7 ;;g 93 n7, 7, 'ps; ggt; ; gy;;; 34.,,3 i,,a; Et e wtin C:rr e.- Cedi', ;-C " N 70 -; ;?r;prict2ry) 2 M i 2 2 ,0000- ^ =
(4vu rium ivi..,), J;nuec, 1075.
- 2. Hargrove, H. G. , "FACTRAN - A Fortran-IV Code for Thermal Transients in a.
UO Fuel Roo", WCAP-7908, June 1972.
- 3. Chelemer H. , Boman, L. H. , Sharp, D. R. , " Improved Thermal Design Pro-
+
cedures" WCAP-8567, July 1975. 4 Burnett, T. W. T., et al., "LOFTRAN Code Description", WCAP-7907-P-A p (Proprietary) and WCAP-7907-A (Non-Proprietary), April,1984. c f*
- 5. " Westinghouse Anticipated Transients Without Trip Analysis" WCAP-8330, -
August 1974. J$i
- 6. ": rite, T , mi oi., "0 cays.m Ruu n,U veuleg, fer 9:;:th " lum lete Trign
?-len;, ', wCAF-15257- A (Fruprietary ) asiu -CA?- 10Z30 A (Non ?reer :t:ry,',
.'un; 1000. i 4i
- 7. U.S.N.R.C. Generic letter 85-05, " inadvertent Boron Dilution Events", .x January 31, 1985. sgr; - >
ey' 8.- !*meiius, T. G. (CJ), " Annuoi ;; pert - 5;;rt "rsj::t, 0 t;t;,, '.000, ~I) E:;t;.;su , 1000 ," : Jet,v N ei... Cvirvi t k.m ;N 1;70, Jun. 1270. (gj
.c
- 9. Liimataninen, R. C. and Testa, F. J., " Studies in TREAT of llrealoy-2-Clad, UO2 -Core Simulated Fuel Elements", ANL-7225, January - June 1966i D. 177, November 1966.
- 10. Bishop, A. A. , Sanburg, R. O and Tong, L. S. , " Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux", ASME 65-HT-31, August 1965.
- 11. Risher, D. N. , Jr. , "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods",
WCAP-7588, Revision I-A, January 1975. Sect M om obc,kd g, i l 15.4-35
-100; Upset; l
- 1. EPRI, "RETRAN-02: A Program for Transient Thermal-Hydraulic Analysic of Complex Fluid Flow Systems", LPRI NP-1850-CCM, Revision 4, November 1988,
- 6. EPRI, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores",
EPRI NP-2513.-CCM-A, Revisior. 2 July 1985.
- 8. " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01", DPC-NE-2004P, Revision 1, February 1990.
- 12. DPC-flE-2011PA, Nuclear Design Methodology for Core Operating Li.T tu of Westinghoase Reactors, Duke Power Company, March 1990.
- 13. Letttr from R.E. Martin (USNRC) to M.S. Tuckman (DPC), " Safety
" Evaluation of Topical Report BAW-10173P, Revision 2, Mark-BW Reload Safety Analysis For Catawba and McGuire," February 20, 1991. D I b
Table 15.4.1-1 (Page 1 of 3) Time Secuence of Events for Incidents which Cause Reacti"ity and Power Distribution Anomalies Accident Event Time (sec.)
. Uncontrolled RCCA Initiation of uncontrolled 0.0 Bank Withdrawal from rod withdrawal from 10 8 of a Subcritical or Low nominal power Power Startup Condition (
Power range high neutron flux 12.1 ll. O g low setpoint reached > Peak nuclear power occurs 12. 7 \l . l ($ Rods begin to fall into core 12.0 ll. S [ ) Peak heat flux occurs n.7 14. 4 ( Minimum DNBR occurs M (3. 5 ( M n , i dt^;- +"" !" 2 - ( occurs ( Peak average fuel temperature 1:. P 13, 5 < occurs Uncontrolled RCCA Bank Witharawal at Power' W 4 @ M- #4J rLei-1, Fa'St reactivity Initiation of uncontrolled RCCA 0.0 insert 4qn - withdrawal at a high reactivity insertion rate (75 pcm/sec)
\Powerrangehighneutronflux 1,55 tpoint reached ht Rods beg all into core 2.05 Mi i d DNBR s 3.20
- 2. Slow reactivit' -Initiation of akQdRCCA 0. 0 insertion withdrawal at a small reacQvity insertion rate (0.6 pcm/sec) 7
/ Overtemperature aT reactor .
71.5 trip signal initiated N Rods begin to fall into core 73.5 Minimum DNBR occurs 5. 0 4 NOG-U
. .. .. m..__.__._- _ _ .- . - _ . . _ . . _ _ _ _ . _ ~ . _ . -._ - _
1 J Initiate Bank Withdrawal 0.0
-Pressurizer Sprays Full On 19.4 Pressurizer PORVs Full Open 21.8 High Pressure Trip Setpoint 36.8 Reached Pressurizer Safety Valves Lift 37.1 Overtemperature AT 37.2 Trip Setpoint Reached Control Rod Insertion _Begins 38.7 l Single RCCA. Withdrawal
-Initiate RCCA Withdrawal 0.0 RCCA Completely Withdrawn 4.2 Pressurizer Sprays Full On 5.2 High Pressurizer Pressure 46.3 i Reactor Trip Sotpoint Reached Control Rod Insertion Begins 48.3 l
l i
. - . , _ . .~ .- . . _ . _ . - , . _ _ , , . _ , - - - - . _ . _ - _ _ . _
Table 15.4.1-1 (Page 2 of 3) Time Sequence of Events for Incidents which Cause Reactivity and Power Distribution Anomalies Accident Event Time (sec.) Startup of an Initiation of pump startup 2.0 Inactive Reactor Coolant Loop at Power reaches P-8 trip 9.1 an Incorrect setpoint Temperature Rods begin to drop 9.6 Minimum ONBR occurs 10.4 CVC5 Malfunction that Results in a Decrease in the n the eac or yN Coolant g Oilution during refueling Dilution begins Operator isoloatet source of >1350 0[/ dilution and minimum margin to criticality occurs
- 2. Dilution ing Dilution begins O startup Operator isolates source o 1620 ilution and minimum may n to c'riticality occurs /
- 3. Dilution during full power operation -
- a. ^utomatic
. One per, cent shutdown margin lost $5421
,eactor control s
- b. Manual Oilution begins 0 reactor contrsf 'Overtemperature ai reactor x'N 53 codrol trip setpoint reached Rods begin to fall into core r5 l
One percent shutdown margin is lost (if dilution continues after trip) Nx
$5421 N
l 49C ', ew a l
la. Dilution-during Dilution begins - power operation- , (manual rod control) Reactor trip setpoint reached 0 Operator terminates dilution <1027 lb . . Dilution during Dilution begins - rower operation ' (automatic rod Rod insertion limit alarm 0 control) s'et point reached Operator terminates dilution <1599
- 2. Dilution during Dilution begins -
startup Reactor trip setpoint reached 0 - Operator terminates dilution <1027
- 3. Dilution during Dilution begins 0 .
refueling High-flux-at-shutdown- 3424 alarm setpoint reached Operator terminates dilution <5224 k s v - e,r ,v.- ., . - - - - -D . .
Table 15.4.3-2 Parameters for Sinole Uncontrolled Rod Withdrawal Dose Analysis
.l.
Data and assumptions used to estimate radioactive sources from postulated accidents,
- a. Power level (MWt) 3565
- b. Percent of fuel defective 1
- c. Steam generator tube leak rate prior to 1 and during accident (gpm)
- d. Offsite power- not available
- e. Initial reactor coolant activity Table 15.0.7-2
- 2. Data and assumptions used to estimate activity released,
- a. Iodine partition factor for steam release 0.01
- b. Release from steam genera *. ors (lbm) 268,550
'(0-30 min) 931,932 (30 min - 8 hrs)
- 3. Dispersion data Distance-to exclusion area boundary (m) 762 a.
Distance to low population zone (m)J 8839 b.
- c. X/0 at exclusion area boundary (sec/m )
(0-2 hrs) 9.0E-04 d, X/O at low population ~ zone (sec/m ) (0-8 hrs) 8.0E-05
- 14. Dose Data
- a. Method of doseccalculations Appendix 15A
- b. Dose conversion assumptions. Appendix 15A
- c. Dose Results (Rem) see Table.15.0.7-1
- - - - - r ,m.e w w ,e w v -- m mJ +,,--,w -v-r -v-- w n e- ey
Table 15.4.8-2 (Page 1 of 2) Parameters for Rod Ejection Accident Oose Analysis
- 1. Data and assumptions used to estimate radioactive source from postulated ac-cidents
- a. Power level (MWt) 3565 So ,
- b. Failed fuel .W percent of fuel rods in core
- c. Activity released to containment atno-sphere from failed fuel and available for release (percent of core gap inventory)
Noble gases 10 Iodines 10
- d. Iodine activity released to containment, 5 sumo and available for release via ECCS leakage outside containment (percent of core gap inventory) ,
- e. Melted fuel None
- f. Iodine fractions (organic, elemental, Regulatory Guide 1.4 and particulate)
- 2. Data and assumptions used to estimate act-ivity released
- a. Containment free volume (ft3 ) 1.038E+06
- b. Containment leak rate (percent of containment volume per day) 0<t<24-hrs 0.3 I ts24 hrs 0.15 l
~
c.' Bypass leakage fraction 0.07 L- d. Offsite power Not available t-k*~ tem e.to, g g.k ng., p g,,, to aec.Qu4 ( i 4eres I
,- , r7 -,-4 ~ .,-
Table 15.4.8-2 (Page 2 of 2) Parameters for Rod Ejection Accident Dose Analysis
- 3. Dispersion data
- a. Distance to exclusion area boundary (m) 762
- b. Distance to low population zene (m) 8850
- c. x/Q at exclusion area boundary (sec/m3 )
0-2 hrs 9.0E-04
- d. X/Q at low population zone (sec/m3 )
0-8 hrs 8.0E-05 8-24 hrs 5.2E-06 1-4 days 1.7E-06 4+ days 3.7E-07
- 4. Dose data
- a. Method of dose calcular. ion Appendix 15A
- b. Dose conversion assumptions Appendix 15A
- c. Doses (Rem) 544 T h IEq1.l Exclusion area boundary Whole body 6.6E-02 Thyroid 8.1 Low population zone Whole body 1.1E-02 t
Thyroid 1.7 l l 12/86
k.Q kace _,, r o'7
/
o g t ks .- e s -
. n . s l Ii ..
j 3
~
11!38{:: , 20.000 - - N(g ,/ 1f' ii \ 1 0 !l1 :: \ I
![bbbb -
85 1,0000 -
\'4
!! MMM :: ,
hg I m . dd5 ..
"T 3 .20000 -
./ 3'g "
(d / [:l "y:8 j g s
\
<!=38bb--
s 02000 - - /
,\*'
t*
\
01000 e eg e e e g- h 8 ' %. 8 8 8 C
*/ . . e W \. c:i d
~
o m o n :- x ~
. .[- \
T1HE
/. (SEC)\.4 j '
\ %
UNCONTROLI.D RCD 'JI""P.DRA'JAL FROM A SUBCRI!!"A1, CONDITICH
.A%u wet McGUIRE NUCLEAR STATION Y Tigure L5.'.1-L
- h. 1986 UPDATE Ad 4
.,ha 1v +-
Seplag t
/
b 4 e i t 0000 J N
\ /
/
N / . 5
.60000'- -
/ --
/ .
L / i C U 5
.60'000 - - -
/ .. 1
. /
x / C
/
d s /
/
y _ . a0000 - - s
/ ..
w i
\
j
- ', /
P b- .20000 - N (J a w -.
/x l' -a v f ,
o f -j x 0.0- e- . e e o e c ' c o o o C C i- C C C Q a ,, o . . . . . e' f D C-ND. O
~
D
~
C m
-; t I
/ TlHE (S,EC)
# 4 I /
s
\.
N, x ,, ilNCONTROLLED RCD ITHOR.WAL FROM A'SUICRITICAL CONDI!!;S l: i psa wm McGUIRE NUCLEAR STATION W r igure 15. \ .1-2 1986 UPDATE x
C@ O ttl ,.
. 3 ,
I (N , I a
- . 4
% s ,' g I
-f,
$ h
. m., ,,-
E 650.00 +
'N,
/ 1 o- -
@ N, m
,/-
v_ sco og 4 e x ...
~,~..~
m
/ _/ ) .'%.
[9 550.00 -
/, / N .,s,, <
,- %.,w.. [
/*^0 00 o o o IN o o o o e o o o o o e o o .
o o o o
- e. o o e o e o
- - ~ ~ -
' NC ($CC) 2500.0 n
!!0.0 -
a y o
~
s 2000lt w ' d :750.0 . N.., _-
'/ <
o ' e . s e ,, '
's, E t500.0 -
g m .,'- e !!!O O t -! ,- ^ \ ' i [ 1000.00 t /' l " '+ c - p - 7,
~. .
- 750.00 '
5 ,, / l
/ 40000 ? o o o e o z r.
p'/ o e o e o O e o e Q o C ..'S
- o. o -
o D o A o i o o - - ~ ~ - l
- t. u r. t. e.. r . i.
!JNCCNTROLLD RCD WITHDPAWAL FRCM A StlBCRI!! CAL CCNDITION
.au w ai McGUIRE NUCLEAR STATION Y Figure 15.. 1-3 e4VV V6 ein i h
. _ . . - .m._. . . . . _ _ _ _ . . _ . . . . . . . _ - ~ _ _ . . _ _ . _ _ _ _ _ . _ . . _ _ . . _ - . _ _ _ . - . _ . _ . . . . _
4 i V,G y PLAc E g _:1.s
=
I.4
].
c c - I.3
~
.=
-o
. -. I .
Du. u-
.g-s D
k .s ~- - m .. ca: , , ac ', ? w ..3- , J - , y 515. S. - s.-- 1.' 2. 4. s. s. Y s. S. is. 1snc secs ,
\/ -
A \
/ -
,V
/ g- I.6
.- / O f x -
.,/ c l.4
-$_ t.2 '
g u 8- \
~8 .
x
\
w .s l
=c
'\
l W .4 a y .2 _. g 8.
- 8. 1. 2 .' 5. 4 5. 6. 7 8. -9 10.
K 11MC iSCCI L , ~.. F UNCONTROLLED RCCA BANK k"ITH-DRAWAL FROM FULL POWER WITH L MINIMUM REACTIVITT TEIDBACK (75 pcm/see WITHDRAWAL RATE)
.ieu rius i McGUIRE NUCLEAR STATION Figure l$ .t. 2 .--_ _____
. . . __ .. . . . , . . _ . , - . . ~ _ < - , . . , - . _ , , _ .
. . -- . - . . . - - . - .. . - ~ - . . , .. . _ . .
r 2GM. - E 2ssa, t 24M.
'M w
*- SM.
=
M h22 ' i e t PIN. 2844.
"/
- 8. l. \2. 5. 4 5.
TINC 6/ (SCCI
- 7. 8. 9. Is.
\
/
/
~ 1900.
w / 5 u /
/
~ 1EM. ,/
E
,.M.
/
.l .
d /
> /
1280.
-/
$/
= 198e. \
I
/~s
$'ees, u
Q. 688. N 8. I. 2. 5. 4. 5. 6. 7 B. %. Is. TINC istcl h UNCONTROLLD RCCA BANK WITH-DRAWAL F20M FULL ?0VER WITH MINIMt'M REACT!v1:Y FEDBACK (75 pcm/sec WITD RAWAL RATE)
~
McGUIRE NUCLEAR STATION
',Q Figure 15.4.2-2
& (-lfl 628.
w
- 618.
w 666. /
'$ / c' N.
u / h 588'. / Y / a . g 518. o
/
Sss. / l'
\, [
558.
- 8. I. 2. \ 5. 4 5. 6. 7 8. 9. ld.
} TINC ISCCi
\ /
\ /
/
/
4 f . 5.5 /
/ -
- 5. l
. / \ 'O /
O 2.5 ,/ s 2. 1.5 l
\
- 8. 1. 2. 5. 4. 5. 6. 7. 8. N IB.
11MC ISCCI l,/ (JMCONTROLLD RCCA BANK WITH-DRAWAL FROM FL*LL POWER WITH MINIML'M REACT!"IIT TECBACK I (75 pcm/ sec WI-*r".JRAWAL RATE) i nu wei W:GUIRE NUCLEAR STATION Y Figure 15..,.2-3 u ur.) Gr uhi c
. - - . . - . - . . . . . = . . . . - - - - . . . . - . ~ . . - ~ . = . . . . - . . - ~ , - . . - -
l.6
;i4 ) ,
b I.2
,,,, g, --
v 5 No N g .6
.. N ,'
s /
.2 %
^!
$l
- 8. =
- 8. 15. 2R. 58. 46. 55. 68 7 t. es.
IlfC ISCCI /
\ /
\
l 7' , 1.6 ! 4.4
/
f l .2 / l -8 1. l : 2 2 /-
- .. / \
5 /
- . 4 / -
.@ ~
/
[- \, 2 i el
/ 8. la. 23. 58. de. 50. 68. 78. St.
i
'/ flMC iSCC8 l
I UNCONTKOLLD RCCA BANK k'ITH-DRAWAL FROM FJLL POWER' k'ITH MINIMUM REACTIVITT FEE BACK (0.6 pcm/see WITHDRAWAL RATE) lM na sses i McGUIRE NUCLEAR STATION T igura .15. 4. 2 -4 1986 UPDATE
3
,f' 1) /l.
2GM. l 5 ,
. - 3 PSM.
b 24M. l vi
\
2SM. l 5 >\ 7 '/ u \
}22es.
d # I 2 8 M. /
\
2000. -
- 8. 18 28. M. 48. 58. S&.' ,d. 98.
\ flMC
\ ISCCI/~
\ /
\
gi . t u
" ISM.
E S y14M. / 5 / A l-
! ,2M.
lt / J i _ l--
- I SM. /
l 5 / 3 '
'\
$ SM. /
}g ./
./
6M.h. IB. 28. 58. 48. 53, 68.- 78. Os. TIMC (SCCI. UNCONTROLLED RCCA BANK WITH- ! DRAWAL FROM FULL POWER WITH MINIM"M REACTIVITY TEEDBACK 0.6 pcm/sec WITHDRAWAL RATET M (McGUIRE NUCLEAR STATIO nu mist l M Figure 15.4.2-5 000 J.".'!I
- 2. . _ _ . _ . _ . _ . . . . . . _ . ._ ._ . .. -.__ ._. _ m _. , _ . - . . _ . - . . . _ . . . . . . _ . . _ . _ _ _ _ . _ . . - .
AtPtAC6 Ste. , GIO.
- 1 3 ,
g 688. kw 54. ,
~ ,
Y,, Sea, j W /
/
w s78. 8 564. 554.
- 8. 18. 29. 58. 45. / 58. ss. 7s. es.
TIE / ISCCI
/
\/
4 5.5
/
5. 2.5 j I-. 2 .!
\
- /l.5 t I.
' E. 18. 26. 58. 40. . 58. 64. 78. 84.
I I tim istcl
!!NCONTROLLED RCCA BANK WITH-i DRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY TEIDBACK (0.6 pcm/sec WITHDRAWAL RATE) has m' . McGUIRE NUCLEAR STATION Y Figure L5.4.2-6 1986 UPDATE .-
-.i
._ _ _ _ _ _ _ . . ~
I l l s {(? 1L/1L :f - l I N 1 i
/ I
\ ! Q
\ x--
'~
~
. - el __' ;]_..'. h. gy l
qq. m eerw;
. . . . _ _ . .sya}- 5g.h.y q
..m (
m# ,. ; m-_ . . ,..
- ..A . . . .
-E'HiEW . ..
=_
.J h ..a. g
,,,, tr - - ~ ~ =:w _= = . m_~3.
.. . _%m~ -
-----.= .
l m-- _ . _ _ _ , - --- ;. ,a
~=-= m ,
i
. ,1yy .,~~ :
b,
.,.w. ro. . .. . A
, r. . . ,..c ... ,.
. , . . , . - .-Pass - . .- - - -wt . ~
+.r 77 . .- - a a,
. y ., 9 t . . . . . . . - . .. I w .: 4 :.., 4 ;.x. y q ;; ; ; a g g . - }i y , ,,, e i G
, : ,.. .e,
' -r t- - .3 . .
H. 41.nh,. 4 w E
-1
~
l s.
> l-D ? . -.. F. R ~g t. .,G k . g 4. Q..; i.. W 5
. c r.,
l3-l -}d M.Q; -{h. ;o.'.- 2
"' i l
===-*%. :. s s 5 : .- .=i= :: ; - a=
==:
l .
- 4. w. : +
.. ...~
22 s
.h. - . '. - . ,
..:-=_._=t..C .--...,,.; N
=
..._ - - , , . =
c a
'"2
_ __ i ;:- l _ . ., -a
.4 2-
. . . gr m- - r e . _t a,.f .
_-r.,,,-..,m.
.o x4 , ,,.
w _2# .. s ,. E n & W u y /-=gi3 9 q: . ve. sas -*' - *
- mah-
,rg
.4 !! ,
RfLE j .. "]Dg . y , M*Mt as w .. . . , w-- W{. g te.4 ng fg.n(} C. g w
.,d % ,,, v. . 33 ,. ,6 a c ,f, w . -
a- < w .. .*-..
. , , .w, . . , . - :c u
~
i f. 4%", ' a.Y*y,- h.i Q,m.m %..__s,. ..,. u..
. - n, kt.%. . }_q. .l.- . .x.
- t. - .. . .
[
. ~.mM._
m .-. .
.. . , _,,f'._. <?)i: , >
~*
_f. 4, - - -".-12;:.=. 4 5 _.::e.
$r g3 .. ...a--
a -- e
~
.. .-.. ",a , ... -
l .. . m _. .. . : ,. . ga-e , 24ExTue l _ ~. l o
/ o cn ce ~ ,o .n , -:
A
- A J J J J J J
! BSNO W WIN!W MINI.YJM DNBR VS. REACTI'l'T" INSERTION RATI (ROD k'I~d-DRAk'AL FROM 100* Pok'ER) pu mm McGUIRE NUCLEAR STATION W mm. ta..,.., ~, G N
9Gl,lC'f 2
- ~*x 3
V . . , . . .
.e: - 3. -
.,.A.. ..
, u .. . .
...,e , . , .,.m 2 . . . ..
u n ,o w..,. m . r. g .-
. - ; . 3 v . .. .
Ld.n:fo9 / s' 9, li \
%..[. m/.ml.4 y l_.j w#p _ .m
- y .u, nw [.
s i: . ; : < :: l u_.,i ; y ; AGl *..- . _
.4.4,sc h.. er-fy M nu[tg @ -? * =m.,s . ~ .
- j r.vs.W -1. c.. gf
-\s . . . . . . w
. . . o
\ =
j* * ,,g =
.g.% 48
\
\ L
.,9
- r. . .. ..; .. .
g;
- . . . ,..mm,..
(g.-
...,f . lr '..' ' '
'*4 4.,'f y' " ? f_.L 1 '. ' *
.. a
- 9. 98
,,, r
/
y ,'J . H.) Y .1-Mif 'd?,fp'+ MR%. O
\ f-N'qq- :- U y .a. #ifd'f :.. . ...-g.w ! : ;/ Gt:'.&V! 2
- p. :
3- f.p p' :. . - ..y .
.:~@
==.w-~ r : =- .
._ g ---- u
= /~
..r ii 1
. -- - i
-. ,/ l g o
- / e
- o B l -
=
< .s
- v o 3 .
- S M-i n,.
- ,=
t w
. ., . y -
l- ...._
-- u o <
a . . = . _ g m:--- -W -
=
_ _ , , .* ,, .,,=a -
~ c.
. .g . m -
x ._
- . =
.Q
+ .. g
- o. .
A P
,q , ' , g..4
. . . . > ,m ,
. . . . .... , s.. -
, 3 cj.: - .
- l. .,if 5 . ..! . ' l, . .. ......s.... '.r.
e,ne.r . .-:; - - r..ag # 4 t8.hm
. o.
. > ~,/
L . . , ., .- .. .m... .... r. _ ~ . . .
.7 . ,
d,f ^ .N g (* k'd.V %M 2 d9 g *, * --
. . ~ . . . . . . .w.. . . .
. w
! ?m .. . .M n .u c . . : w y , + : = s.. Et
" . . _ _.=wa m e v g m..g w .ay w o
- _ 3 t r
3 e. s. m
-- .,,_,,,,,,.g,
. u-s
.,. . o.
I; l:y - e: . q+ w: '. /$- ' y y- g -
===*-m=='* T====a===*
ampua a _ . ' ., a ' . . .. - ! mc - v- . -+- ~~. -'.._-~~;... w - 3 .: g ._ : .@ , t . -
- - ' : -
- g-
- y. . . . _ s . . w . .. . .- . . . . < . .,_
I. .. . , ,. . ;.:m..
.:. ,, :. o > - : I y.,- , .
.mww _ .;,
a . . <.a.. , ..
.-- - 1 -
# .m o
- : e. L .e w - .. _ e- ~
.w 5 , semamapp. : w.
. . . _ . . . . .w -
t, b ? .:; . :+ , v. r-- :_ . _ _ . .. . . _ _ .
- l. -., . . . . . . . ..g. . . . , . . . .
b i - 1; ??.il.' T ,'\ 7MS - -f .. 'Y.em man - srb - . - ar
. g2-,3 . . ans as -+
-. -w'. man-==a==
i
--. - - - = , ~
l __ o t L __ .. .. . 1 .- o i q/. 9 --
- c. 9 \- 9
. t n ~ ~ ~
/ i%G WIW!k!W' MINI.WM DNBR VS. REACTIVI"'T INSERTION RATE (ROD '='ITH-DRAWAL l' ROM 60*., POWER) nm McGUIRE NUCLEAR STATION \
Y Tigure 15.4.2-8
\
1986 UPDATE
\ \
\
\
--__ _.= --- - - -- - - ___ ___ - __m__ _ _ _ * - '-- - - . - .. - er ---e- - - - . - rq .v- p ---y e w- m.r r- te'-- % g- e.-.+-----T-e
lll I l ;
!1 .
; Il' i
3%\ h$ s f o
\
?c
[ ( p' ,; . j* i
!- ;} t tI _
r m"!! ; ,
,l LqI i, tff, : j 1!
il, n 0 I .
. 0 5
.L . L 1[ i I
llIl } Ih0 i ,
- l. r l
. 0 2
c m i L=a - 0 0 1 s" .
)
o 0 m! s 1 u (i X C E 5
, n q .!
0
/
K
,i , 1 ,
2 J (
; [
i A t - o t R - N O I _ T R r 0 E t S
- p -
5 N I T T Y
- D -
T
, j!
s ! n T O
>0 I
V I I 2 C
", A
. .,.m.
[ R N 0 - . , 1 i
.. q ; ' .
5 0
- !lu j! L l
,j ,
2
' l 0 L{! ,
l 1 o. s z 2 , l ga i/ j!s2!: j! { meg'" rN -
)g d. hm!
! j
- I; UN lE" h, ,e g !gm 2gE> .
. a d'"
h s-
; l l
. 1 , i':
E'SAR.SECTION 15%.1 - BANK WITHDRAWAL AT!HZP IM see -- -
/
/ .
./
2 i 9 Y I ; t i 9
~
?
=
l
-ism ,
- +. . - . . . . . . . . .
12 34 16 18 29 6 8 le , o 2 4 ! TIIE (SECD@SI i
FSAR SECTION 15.4.1 - BANK WITHDRAWAL AT HZP s b
-M ?-
i &
- E a a '
is. p
" ?
T M
- n. _,
tee ne W
\
t 0 2 4 6 0 le 12 14 16 18 29 TIfE (MIX 3058 i s 4
a FSAR SECTION 15.4.1 - BANK WITHDRAWAL AT HZP 1. 49 b m = e.a
\
e7 n-, 3 1 e.s $ i g
/ -
- ..s x P' i
f \ $ 3 ... - 4 I,t y e.3 N :
- s.2 .
W 3 a.: ; aa , . . . . . . . . . . ; 6 8 18 12 14 16 18 29 O 2 4 Tite (!EEDeSi m _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ . _
FSAR SECTION 15.4.1 - BANK WITHDRAWAL AT.HZP , 9 I 11ee r
\
3 io 3. w 0 i-- Q, = = ! y < C . () $ O e 5-a 5
]
M . , . . . . . . i = g 0 '2 4 6 9 le 12 14 16 "R 2e TIPE tSEIXN2)
ms FSAR SECTION 15.4.1 - BANK WITHDRAWAL AT HZP GS8 648 5 630 2 3 s2e 0 & u 618
\
[c ' 5 \
?*
b 598 l Y'
. j Ut o
w 55" /
)
SGB I / (
~
sse e 2 4 6 8 18 12 14 16 38 22 , TIPE (5EIGGS3 1
4 b 7SAR SECTIOX15.4.1 - BANK WITHDRAWAL AT HZP 1 5 1 4
/ ;
{
\- 3 3 93
- s
$ N
' f e- , , , , , , , , , , , a 2 4 6 8 IS 12 14 16 IC 2R TIE (EEDSS) I u k
Figure 15.4.2-1 7: M Ed .
>e C l k
8 o _. k e U J ? g
- D ol $
5: .- m 3 16
~ ,
5 N F u CD N~ = Cn d ! g-en I.L - o g g s e a hvN!VJON 30 %) 83 Mod HOB 103N l- -
. . . J
FSAR SECTION 15.4.2 UCBW @ 10 % POWER 1.2 l r 7 1.0 N 5 O
=
0 08 O E. T; a 55' l 4 x - 0.6
]
- y .
f I to 0 - - g 04 j
<t w !
8 O O.2
- 0. 0 0 20 40 60 ,80 TIME (SECONDS) ,
.! ;f, Y ' . I ,fi[itf1 ' ,h fI fir , ,l .fI> f! ; !! ?p [!
U t a e. wCnO 0 8 R E _ W _. O _. - P 0
% 6 -
0 1 W - B ) C _ S D N U _ 0 4 O C E 2 S ( 4 E M I 5 T 1 . N O I T C _ i I 0 2 E _ S R A . S F i1
- 0 0 o 0 O
O 9 5 e 7 5 g G r g 0mb mE3<EmbgM .-
<m
- aOO3-
, I sI . 'i ?- li? ;4 e' i ;
!4 :i- : 1 . 2!i ;
<.)I-
- . 9!tht Iil{!
; ) ![. P; !rI It!;(* .[;' 7t!, !! i![![ ! ,t r!-
- C O Wu.A.u*&
0
. _ 8 R -
E W _ O P _ 0
% 6 0
1 W B ) S C _ 0 D N U . _ 4 O C E _ S 2 . ( E 4 M I T 5 1 - N O I
~
T - 0 C 2 E S _ R _ A - S . F .
~
0 0 o
)
x 0 o s 2 ( 4 a L 8 1 i c5L-6 g5 U 2 5~CDmasEe u
< .; ;I , ! ;ie, . , , ' :! 4 ;i , ..!i ,
a ,{'
Figure 15.4.2 5 s e l
.Y
' 4 9y mm s g ..'%.. <
/
h, , < e=== d e q v#88.4 ..., W **%., m %..,
'e, k
%di
- V
~/ /
m ,,.* D . **
- ... _ a ~
a r1 . g y . . a in ....- s-
~
i 2 O
, 1 i
M i ,a a 5 l l . i M l 1, . e 4
- ;: h a 1 'A y 5 5
1 .
.v o 5 :
A*s
'A, M l I
e i f f ., e a d atJO l'Jnn'Jillin'd
Figure 15.4.2 6 e i C a
-1 I 4 l 1
H. - ty < p
.O >
i, f% :
- / t o
e m - l , ! n l w ,
' In o l
a /
- I
-QI l
j O ,
! u V, ,-
/ W
~~
# * .1 e
3 N. , ui T.
- 10 $
- 3 i
-z . J e
v 4 IP k rs G , 2
=
- -*r1 g ,
- 7) [ e
- 0 4-I l Cn, r , : ,
l l l l m
\A
~;
l I l-l 'd8NO WOWINIW i- ,
.--.,,,,,.w...-rv..-, ...,f..,- ..,,-,# ...,.n--, , . .=,.r v.~,.-,-,- ,.-er, , , . . , - - , , - , , , . ,m.__, --.ew,-,.,,,ow--4i-.e-.-,.,-,r- .--.,r-. ...mc....,----,a
1 1 Fleure 15.4.2 7 ' o t o. i y , 1 < i t O c ' . C C h_ h_ r
- 5
~
= = .
I M s I u Q o
] . "
u Ol . . ui T. - <
- r 10 ,
I Z ' O re su b U , r.c Co < W l f l Cn . ,:
-r. -- '
l l - 1, c
'.n l "
I i 88t40 i'JOWitJIW l -. o 4,e--a ve-,v, .-e,w- g -,. _a cm..ew-rw es e w-c y %- v-.sw+gt-se a -=w. g.-m, e. r mg---g,ei -ww a9 iiirp---. WtrY.-- 9 y m w, g reip wr* y +'-gt-7 B'? WTr"- e ntts-m-Wmw-g*'*
FSAR SECTION 15.4.3d SINGLE ROD WITHDRAWAL 120 __ k - ..- - _. 300 . _ . _ _ _ _ . . _ . _
, ~ ^ - ~ ~ - ' ' - - ~ ~
~ ~ ~ ~ ~ ' ~
80 - 3 O 3 7: m 0 C
~
M y . O u 8 40 E 20 K N 60 80 20 40 O TIME (SECONDS)
g & w . i FSAR SECTION 15.4.3d SINGLE ROD WITHDRAWAL 12 l 1 l i 10
- 7 3
! O , = ts. l 08 i
.- M I. O n
,t 4 = i E-C K ,
> 06 m .us
.s
>- Y
- 5 I
i m
. . - - . . . - - . ~ - . - . , g .g 4
E i
; m
!. E o
- o - --
Oi l 4
^
00 - 60 80 G 20 40 TIME (SECONDS) i )-
1 1 i FSAR SECTION 15.4.3d SINGLE ROD WITHDRAWAL ' i G10 i c ~ i ' O O
- i i
tu tr 3 J.f i
@ l Q
! b 3 590 ;
" .m i
.M
- j o ~
N
- n' ;
4 O 9 m I 580 T f t 570 80 0 20 40 60 8 TIME (SECONDS) I
J k-FSAR SECTION 15.4.3d SINGLE ROD WITHDRAWAL 2500 i 2400 5i 2300 C E E 2 i a 2200 ts m Y a 8 m g 2:00 2000 1900 60 80 0 20 40 TIME (SECONDS)
15.6 DECREASE IN REACTOR COOLANT INVENTORY A number of events have been postulated which could result in a decrease in Reactor Coolant System inventory. Detailed analyses are presented for several sucn events which have been identified as limiting cases. Discussions of the following events are presented in this section:
- 1. Inadvertent opening of a pressurizer safety or relief valve.
2 Besak. W *iNstewm s eit On e. oc .thse line s from reeelor c.oolord, pressure. M,y 3/ Steam Generator Tube Rmture. M Pt astrok c.o+1oinmerA. M BWR Piping failure outside containment (Not applicable to McGuirg). 5'/ Loss-of-Coolant Accident resulting from a spectrum of postulated piping breaks within the Reactor Coolant Pressure Boundary. 6 ,/. A number of BWR transients (Not applicable to McGuire). eta'ils bd.h CuJrben IIr oed E e.wsAs Items 1 and 2 are considered to be AN5 Conoitton 11 events, item
- nridue; i: M u. "% C:nciti;c ::: n :rt, and item 3 is considered to De an ANS Condition IV event. Section 15.0.1 contains a discussion of ANS
)
classifications and applicable acceptance criteria. {; 15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY OR RELIEF VALVE 15.6.1.1 Identification of Causes and Accident Descripti g An accidental depressurization of the ReactoT Coolant System could occur as a result of an inadvertent opening of a pressurizer relief or safety valve. Since a safety valve is sized to relieve approximately twice the steam flowrate of a relief valve, and will theNfore allow a much more rapid depressurization upon opening, the most severe core conditions resulting from an accidental depressurization of the Reactor Coolant System are associated with an inadver-tent opening of a pressurizer safety valve. Initially the event results in a rapidly decreasing Reactor Coolant System pressure until this pressure reaches a value corresponding to the hot leg saturation pressure. At this time, the pressure decrease is slowed considerably. The pressure continues to decrease througnout the transient. The effect of the pressure decrease would be to
- decrease power via the moderator density feedback, but the rod control system l (if in the automatic mode) functions to maintain the power and average coolant temperature until reactor trip occurs. Pressurizer level increases initially l due to ext.ansion caused by depressurization and then decreases following reactor trip.
The reactor may be tripped by the following Reactor Protection System signals:
- 1. Overtemperature AT l
- 2. Pressurizer low pressure An inadvertent opening of a pressurizer safety valve is classified as an ANS Condition II event, a fault of moderate frequency. See Section 15.0.1 for a discussion of Condition 11 events.
, 15.6-1 l 1986 Update
15.6. 1.3 Conclusions
%e results of tne analysis snow that the pressurizer low pressure and the
- <ertemperature 2T Reactor Protection System signals provide accouate protec-tion against the RCS cepressurt:stion event. No fuel or clad damage is pre-g j icted for this accident.
ha 15.6.) STEAM GENERATOR TUBE FAILURE
% Lad Poc3e,4 15.6,/.1 2
Identification of Causes and Accident Descriction The accident examined is the complete severance of a single steam generator tuce. The acciaent is assumed to take place at power with tre reactor coolant contaminated with fission proaucts corresponding to continuous operation with a limited amount of defective fuel rods. The, accident leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power, or failure of the condenser steam dumo system, discharge of activity to tne atmosphere takes place via the steam generator safety and/or power operated relief valves. In view of the fact that the steam generator tube material is Inconel-600, nich is highly ductile, it is considered that the assumption of a complete severance is somewhat conservative. The more probable moce of tube failure
.ould be one or more minor leaks of undetermined origin. Activity in the Steam and Power Conversion System is suoject to continual surveillance and an accumu-lation of minor leaks wnich exceeds the limit established in the Tecnnical Soecifications is not permitted during the unit operation.
The coerator is expected to determine that a steam generator tubt rupture has occurred, and to identify and isolate the af fected steam generator on a res-tricted time scale in order to minimize contamination of the seconaary system and ensure termination of racioactive release to the atmosonere f rom the affected unit, The recovery proceaure can be carriea out on a time scale which ensures that break flow to the seconaary system is terminatea before water ievel in the affected steam generator rises into the main steam ciping. Suffi-l c'ent indications and controls are provided to enable the coerator to carry out tnese functions satisf actorily.
;mmeolately apparent symptoms of a tube rupture accident sucn as f alling
- ressurizer pressure and level and increasea cnarging pump flow are also symptoms of small steam line breaks and loss of coolant accidents. It is trerefore important for the operator to determine that the accicent is a j uoture of a steam generator tuce in orcer that ne may carry out ne correct ecovery proceaure. The accicent uncer discussion can ce icentifdea oy the I following method. In the event of a complete tube rupture, the reactor coolant l system pressure decreases and the concenser air ejector raaiation ana/or steam generator blowoown radiation monitors exnibit abnormally high readings. If the c:ntainment cressure, containment radiation and containment recirculation sumo evel exhibit normal readings, inen a steam generator rupture is diagnosed to rave occurred.
l C:nsideration of the indications provided at the control board, together with tne magnitude of the break flo., leads to the conclusion tnat the accicent
.5.6-3 L984 Locate
(
..._.____ ~ _ _ _ _ _ . . _ _ _ _ _ _ - _ _ _ _ _ _ _
1 15.6.2 BREAK IN INSTRUMENT LINE OR OTHER LINES FROM REACTOR CCOLANT PRESSURE BOUNDARY THAT FENETRATE CONTAINMENT 15.6.2.1 !centification of Causes and Accident Descriotion Instrument lines connected to the RCS that penetrate the containment are discussac in Section 6.2.4. There are also the sample lines from the not legs of react:r coolant locos A and 0, ne samole lines from the steam anc licuid space of the cressurizer, and tne Cnemical anc volume Control System (CVCS) letdown and excess letdown lines that penetrate the containmer,t. The sample lines are proviceo with normally closed isolation valves on cotn sices of the normally open containment isolation valves on cotn sices of the containment ; wall. In all cases the containment isolation provisions are designeo in accorcance with the requirements of General Design Criteria 55 of Appendix A to
- 10 CFR 50.
15.6.2.2 Analysis of Effects and Consecuences The most severe pipe rupture wita regard to radioactivity release during normal plant operation occurs in the CVCS. This would be a copolete severance at rated power conaitions of the 3 inen letdown line just outsian containment, between tne outeoard letdown isolation valve and the letdown neat exenanger (see Figure 9.3.4-1). The occurrence of a comolete severance of the letcown line would result in a loss of reactor coolant at the rate of acoroximately 140 gps (referenced at a censity of 62 lb/fts). Since the release rate is within the capability of the reactor makeuo system, it would not result in Engineered Safety Features System actuation. Area radiation and leakage detection instrumentation provide the primary means for detection of a letdown line-rupture. Frecuent operation of the CVCS reactor makeup control system and other CVCS instrumentation would aid the operator to identify and isolate the rupture within 30 minutes. Once the rupture is identified, the coerator would isolate tne letdown line rupture ey closing the letdown orifice isolation valves and the pressurizer low level letdown isolation valves. Alternatively, the operator would close the letdown line isolation valve outside containment to isolate the rupture. All valves are provided with control switenes at the main control board. There are no single failures that would prevent isolation of tne letdown line rupture. The rupture outside containment of any other small line connected to the reactor coolant system can be isolated and will have less severe consecuences
<with regard to release of reactor coolant.
(- inree cases are analyzed: Case 1: Normal eouilibrium Tecnnical Specification iodine concentrations exist at the timoref the accicent. Case 2; inere is a pre-existing icdine spike at the time tne accident occurs. The reacter coolant concentrations are tne maximum cermitted for full :ower coeration. Case 3: Tnere is a coincicent iocine soike at tne time the accident occurs. The iodine concentrations are found by incress1ng the eauilibrium appearance rate in the coolant ey 500, Otner sssumotions anc carameters are f ounc in Table 15.6,2-1.
~ _ _ . ._._ . _ _ _ _ . . . _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . . _ _ _ _ . _ . .
15.6.2.3 Envir nmental consecuances Based on tne foregoing mocel, the coses to the thyroid and whole booy are calculateo at exclusion area bounoary and the low population zone. The results appear in Table 15.6.2-1. Therefore, the radiological consecuences resulting from failure of the 3 inen CVCS letdown line is not expectea to exceec a small fraction of the 10 CFR 100 cose limits.
. . . - _ - - - - _ - - - . ~ - . - ~ _ . . - . - - . -
l l l A steam generator tube failure is classified as an ANS Condition IV event, a i limiting fault. See Section 15.0.1 for a discussion of Condition IV events, l 3 , 15.6.7.2 Analysis of Effects and Consecuences l Method of Analysis Ier t due. CMcMon prlmes InestimatingthemasstransferfromtheRCSthroughthebrokentubekthe following assumptions are made:
- 1. Reactor trip occurs automatically as a result of low pressurizer pressure, ,
- 2. Following the initiation of the safety injection signal, two centrifugal pumps are actuated and are assumed in the analyses to continue to deliver flow for 30 minutes. ,
- 3. Af ter reactor _ trip break flow reaches equilibrium when it is balanced by 3
incoming safety injection flow as shown in Figure 15.6.2-1. The resultant break flow is assumed to persist from plant trip until 30 minutes.after the accident. No operator act1uns are assumed to reduce break flow. 4 -The steam generators are_ controlled at the safety valve setting rather ' than the power operated relief valve setting.
- 5. The operator identifies the accident type and terminates steam relief from the faulted steam generator within 30 minutes of accident initistion.
The above assus.ptions, suitably conservative for the design basis tube rupture, are made to maximize doses and do not explicitly model operator actions for covery. 3 4 Results M Figure 15.6.f-1illustratestheflowratethatwouldresultthroughtherup-tured steam generator tube. The previous assumptions lead to a conservative upper limit estimate of 125,000 pounds for the total a:1ount of reactor coolant transferred to the secondary side of the ruptured e ~4m oenerator as a result of a tube rupture accident. 3 15.6./.3 Environmental Consecuences
.The postulated accidents involving release of steam from the secondary system
- do not result in_a_ release of radioactivity un_less there is leakage from the RCS to the secondary system in the steam generators. A conservative analysis of-the postulated steam generator tube rupture assumes the loss of offsite j' power. This causes the loss of main steam dump capabilities.and.the subsequent l venting of steam from the secondary system to the atmosphere. A conservative l analysis of the potential _offsite' doses resulting frem this accident is present-l ed assuming primary to secondary leakage. This analysis incorporates assumptions of I re ce" de'ect M fel and steam generator leakage of'1 gpm prior to the peltulated accident for_a time sufficient to establish equilibrium specific ac tivities in the secondary system.
N M4CMcobos tohe, c.oncord.rahih 15.6-5 ;984 Update
_ _ _ _ _ . ~ . _ _ _ _ _ 1 The ONBR calculation for this accident is performed with the VIPRE-01 computer code (Reference 18) using the Statistical Core Design procedure described in Reference 19. DNBR is a concern for this j transient because the assumed loss of offsite power causes a reactor coolant pump coastdown. Because of the loss of inventory through the ruptured tube, the RCS pressure is significantly lower than normal j when the coastdown occurs. Since the loss of offsite power is assumed to occur coincident with reactor and turbine trip, the amount of depressurization prior to the coastdown would be limited by the , i overtemperature AT trip function. This trip setpoint is reduced both by depressurizations and RCS heatup. Because of the relative effects . on DNBR of the heatup and depressurization allowed by this trip - function, the steam generator tube rupture coastdown transient from a depressed RCS pressure is bounded by the feedwater line break coastdown transient from an olevated RCS temperature. DNBR is therefore not explicitly calculated for the tube rupture transient, but the result is bounded by the calculation for feedwater line break
-in Section 15.2.8.2.
P f 5 L l t l i 4.._ . . . . , - , . , . . . . . ., . . - , . _...,,-_,,_..-_m.,_..m.,.,_,,..__...-,,-_,,_.,c._.+,.-_, .,,....._--7 :___...,
i The following assumptions ana carameters are useo to calculate the activity release and offsite dose for the postulateo steam generator tube rupture: i
- 1. Prior to.the accident, an eouilibrium activity of fission products exists in the primary ano seconoary system caused by primary to secondary leakage '
in the steam generators.
- 2. The accident is initiated by the rupture of a steam generator tube which results in the transfer of approximately 125,000 pounds of reactor coolant into the shell side of the cafective steam generator during the first 30 minutes.
- 3. Offsite power is lost at the time the accident occurs.
4 The primary to secondary leakage in the three nondefective steam ' generators is 0.25 gpm per generator from 0-2 hours and 0.333 gpm per ~ generator from 2-8 hours.
- 5. The steam release from the defective steam generator terminates in 2 hours. The release from the noncefective steam generators terminates in 8 hours.
- 6. All noble gases which leak to the secondary side are released. ,
el
- 7. The steam generator iodine partition factor is 0./ during the accident.
- 8. Water density at 180'F is used to convert condenser volumetric flow to mass flow.
- 9. The steam generator blowoown rate is 50 gps.
- 10. The turbine building steam leak rate is 1700 lbm/hr.
- 11. The condensate leat rate is 7510 lbm/hr.
l 12. The glana seal iodine partition factor and the condenser air ejector iodine partition factor are each 0.15. 3
- 13. Other assumptiens are listed in Table 15.6.A-1.
Based on the model in Appendix 15A . the thyroid and wnote bocy cosas are l calculated at the exclusion area councary and the low population zone. The ! results are cresented in Table 15.6. P 1. The coses are within the limits of 10 CFR 100. J 4 15.6 7 SPECTRUM OF EWR STEAM SYSTEM PIPING FAILURES OUTSIDE CONTAINMENT This Section is not applicable to McGuire. . G l-l 15.6-6 1986 Update
l l empirical FLECHT correlation is replaced by the BART code. BART employs rigor- j ous mechanistic models to generate heat transfer coefficients appropriate to the actual flow and heat transfer regimes experienced by the LOCTA fuel rods. This is consiuered a more dynamic realistic approacn than relying on a static empirical correlation. An error in BART modeling of the thimble tubes and hot assembly bundle power . was discovered after this ana'ysis was performed. The Westinghouse evaluation f of this error is addressed in Reference 17, Le,t Abebe 1 Small Break LOCA Evaluation Model The NOTRUMP computer code is used in the analysis of loss-of-coolant accidents due to small breaks in the reactor coolant system. The NOTRUMP computer code is a state-of-the-art one dimensional general network code consisting of a number of advanced features. Among these features are the calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drif t flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, and regime-dependent heat transfer correlations. The NOTRUMP small break LOCA emergency core cooling system (ECCS) evaluation model was developed to deterrine the RCS response to design basis small break LOCAs and to address the NRC concerns expressed in NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants." In NOTRUMP, the RCS is nodalized into volumes interconnected by flowpaths. The broken loop is rndeled explicitly with the intact loops lumped into a second loop. The transient benavior of the system is determined from the governing conservation equations of mass, energy and momentum applieo throughout the system. A detailed description of NOTRUMP is given in References 11 and 15. The use of NOTRUMP in the analysis involves, among other things, the represen-tation of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation. The multinode capability of the program enables an explicit and detailed spatial representa-tion of various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant transient. . Cladding thermal analyses are performed with the LOCIA-IV (Reference 8) code which uses the RCS pressure, fuel rod power history, steam flow past the uncovered ptrt of tne core, and mixture height history from the NOTRUMP hydraulic calculations, as input. l A schematic representation of the computer code interfaces is given in Figure 15.6.4-3. l The small break analysis was performed with the approved Westinghouse ECCS Small Break Evaluation Model (References 8,11, and 15). Large Break Input Parameters and Initial Conditions Table 15.6.4-1 lists important input parameters and initial conditions used in the large break analyses. 6'h 6 tk m -L & 15.6-12 12/87 f 1
i Attachment 1: ! Post-LOCA Suberit icality Fvaluation l An analysis has been performed to determine the sump mixed mean boron concentration as a function of pre-trip Reactor Coolant Systen. (RCS) ! boron concentration for a postulated large break Loss of Coolant Accident (LOCA). Water mass contributions from the RCS, Cold Leg Accumulators (CLAs), Refueling Water ftorage Tank (FWST), Ice Condenser, and Emergency Core Cooling System (ECCS) and containmett spray piping have been taken into account. This analysis used the principle input parameters provided in Table 6.3.3-3. High concentration borated water volumes (e.g., FWST and CLAs) are -! conservatively minimized using Technical Specification minimum al2 owed 7 values minus associated measurement uncertainties. A berated water !
- mass contribution from Ice Condenser melt has been evaluated from the start of the LOCA until the initiation of cump recirculation.
Potential borated water holdup in upper containment from the , t initiation of normal containment spray was taken into account. > Results of the analysis are compared with the required boron concentrations necessary to keep the core subcritical, with no credit taken for control rod insertion, during the sump recirculation mode. The analysis provides an available sump mixed mean boron concentration curve that_must bound the required all rods out (ARO) critical boron cencentrations for each cycle. The required ARO critical boron concentrations are evaluated for each core design as part of the reload safety analysis process. Since much of-the ECCS piping is used during normal operation for residual heat removal and normal chemical and volume control, much of this ECCS piping contains relatively low concentration borated water. A single failure of an ECCS train would therefore be non-conservative, I and was thus not considered in the Post-LOCA Suberiticality Analysis. Attachment 2: Table-15.6.5-1A lists input parameters for the post-LOCA suberiticality analysis. i A- - ,,-sm.m .,w.w.,<ne--->a caww- -,,,,N-r ,-,,,,-,.,vwv. - em-ows,m,, ,- ,,,,.e - uw s: e.- & ws , . . w m m ., w m- m w -e--e -s w g w u -,,--- -<n,,-- - ,
transfer enhancement predicted during reflood. The various fuel assembly specific transition core analyses performed resulted in peak clad temperature increases of up to 10*F for core axial elevations where PCTs can possioly occur. Therefore, the maximum PCT penalty possible for 17 x 17 0FA during transition cores is 10'. Once a full core of the 17 x 17 0FA fuel is achieved, the large creat LOCA analysis with UH1 removed will apply without the crossflow penalty. C 15.6./.3 Environmental Consecuences The postulated consequences of a LOCA are calculated for 1) offsite and 2) control room operators. Offsite 00se Consecuences The offsite radiological consequences of a LOCA are calculated based on the following assumptions and parameters.
- 1. 100 percent of the core noble gases and 25 percent of the core iodines are released to the containment atmosphere.
- 2. 50 percent of the core iodines are released to the containment sumo and available for release via ECCS leakace outside containment.
- 3. Annulus activity which is exhausted prior to the time at which the annulus reaches a negative pressure of -0.25 in. w.g. is unfiltered.
- 4. ECCS leakage begins at the earliest possible time sump recirculation can begin.
- 5. ECCS leakage occurs at twice the maximum operational leakage.
- 6. Bypass leakage is 7 percent of total containment leakage.
- 7. The effective annulus volume is 50 percent of the actual volume.
B. The annulus filters become fouleo at 900 seconds resulting in a 15 percent reduction in flow.
- 9. Elemental iocine removal by the ice condenser begins at 600 seconai and continues for Anap seconas with a removal ef ficiency of 30 percent.
3S61 10, One of the containment air return fans is assumed to fail.
- 11. The containment leak rate is fifty percent of the Technical Specification limit after 1 day, of
- 12. Iodine partition f actor for ECCS leakage is 0.1 for the course of the accident.
- 13. No credit is taken for the auxiliary building filters for ECCS leakage.
- 14. The recundant hydrogen recombiners and igniters fail. Therefore, purges are recuired for hydrogen contro',
15.6-17 M2L
l
- 15. The annilius reacnes equilibrium after 200,000 seconds sucn that the only discnarga 15 aue to inleakage.
- 16. Water density at 160*F is used to calculate the sump water mass.
- 17. Elemental and particulate iodine removal f rom the containment atmosonere by containment spray begins at the initiation of the event and continues -1 and f or 1/2 minutes with spray removal lambaas of 1.52E-02 minutes 4.10E-02 minutes- 1, respectively.
- 18. Other assumption are listed in Table 15.6.4-9.
Based on the model in Appenaix 15A, the thyroid and wnole body doses are The calculated at the exclusion area boundary and the low population zone. doses are presented in Table 15.6.5-9 and are within the limits of 10 CFR 100. Control Room Ocerator Oose The maximum postulated dose to a control room operator is determined based on the releases of a Design Basis Accident. In aadition to the parameters and assumptions listed above, the following apply:
- 1. The control room pressurization rate is 1.000 cfmt the filtered recircula-tion rate is Q cfm.
- 2. The unfiltered inleakage into the control room is 10 cfm.
"hd k. f 05. M, S ".'hed. .E* W _ , , , . _ _ , _ _ _ m , _
.m..
4 15.6.y A NUMBER OF BWR TRANSIENTS Not applicable to McGuire. 12/?8 15.6-18
- 3. The VC fan rated flow rate is 1800 cfm.
- 4. One of the two filtered intakes is assumed inoperable for the first four hours of the accident; i.e., the 1-4 hour X/O value is doubled.
L. Other assumptions are listed in Table 15.6.4-10. Based on the model presented in Appendix 15A, the thyroid, skin and whole body doses are calculat ed in the control roem. The doses are presented in Table 15.6.4-10 and are within the limits of 10 CPR 100. I
. .. . _ . _ - _ - - _ . ~ _ _ .
REFERENCES FOR SECTION 15.6 (cont':)
- 16. Rahe, E. P. (Westinghouse), letter to Tedesco, R. L. (USNRC),
No. NS-EPR-2538 December 1981.
- 17. Young, M., et. al., "BART-1A: A Computer Code for the Best Estimate Analysis of Reflood Transients", WCAP-9561-NP-A, Addendum 3, (Non-Proprietary), 1986.
Insert from edoc.kscl page. l 15.6-20 12/88
i f r
- 18. EPRI, "VIPRE-01: A Thermal-Hydraulic Code f or Reactor Cores',',
FPR7 NP-2511-CCM-A, Revision 2, July 1985,
- 19. " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01", DPC-NE-2004P, ;
Revision 1,-February 1990. i l t I I w' n + w r , -r w a ~,,,---,e.,rw,-.---w--- a+ , - - + - - ,--r-, w-- - . , - , ,.--a-n. - - - , ,,ww.en-e-----a,- ---ow- 4.,-rw<o -
, , 4 s n.-sn m-- -s,--
. _ _ - - _ _ _ . - . . . _ _ - . _ _ _ _ _ _ . _ _ __ _ ._____.__..._m ._.. _ ,
{ I Table 15.6.2-1 : Parameters for Postulated Instrument Line Break Accident Analysit P 0;n ;; c . ; t i '. : - Re:!' tic !
-1. Data and assumotions used to estimate
- 2;;listi; :;ee radioactive source from postulateo r:t ;n;i i J.
accidents
- a. Power level (MWt) 3565.
- b. Percent of fuel defected 1.
- 2. Data and assumptions used to estimate act- '
tivity released of
- a. Iodine partition factor 0.7 too
- b. Flow rate through ruptured line (gpm) ,140.
- c. Time to isolate break (min.) 30. *
- 3. Dispersion data ,
- a. Distance to exclusion area boundary (m) 762.
ssso-
- b. Distance to low population zone (m) . 6095.
4.o
- c. x/Q at exclusion area boundary (sec/m3) Gr6 E-4 8.o
- d. x/Q at low population Zone (sec/m3) tr6E-5 A. Oose cata
- a. Methoo of dose calculations Regulatory Guide 1.4
- b. Dose' conversion assumotions Regulatory Guides L 1.4 and 1.109
- c. Ooses (Rem) set. T A w is.o 7.I i
D555 Case-1-(No'iodinesoike) Exclusion Area Bounoary < Whole Body 1.6E-1 Thyroid 3.2E-1 Low Population Zone Whole-Body 5.1E-3
.hyroid 1.0E-2 l
i-L _ . _ ~ - _ _ _ _ -. _ _ . _ ~
i 3 Table 15.6 7-1 (Page 1 of 2) Parameters for Steam Generator Tube Ruoture Dose Analysis
- 1. Data and assumptions used to estimate radioactive source from postulated accidents 3565
- a. Power level (MWt) 1
- b. Percent of fuel defected 1
- c. Steam generator tube leak rate prior to accident (gpm)
- d. Offsite power Not available
- e. Reactor coolant activity Table 15.0.8-2
- 2. Data and assumptions used to estimate activity released of
- a. Iodine partition factor during 0.)'
accident 50,000
- b. Steam release from defective steam generator (1bm)
- c. Steam release from three r.ondefective steam generators (1bm) 412,000 (0-2 hr) 978,000 (2-8 hr) 125,000
- d. Reactor coolant released to the defective steam generator (1bm) l
- 3. Dispersion data 762 l a. Distance to exclusion area bouncary (m)
Distance to Icw population zone (m) 8850 b. E96- sp.ase
i 3 Table 15.6./-1 (Page 2 of 2) Parameters for Steam Generator Tube Rupture Oose Analysis
- c. x/Q at exclusion area boundary (sec/m2) 9.0E-04 (0-2 hr)
- d. x/Q at low population zone (sec/m2) 8.0E-05 (0-8 hr)
- 4. Dose data
- a. Method of dose calculations Appendix 15A
- b. Oose conversion assumptions Appendix 15A
- c. Doses (Rem) See. Tc h l5.0.*l.1 Exclusion area boundary l Whole body 2.8E-01 Thyroid 4.1 Low population zone Whole body 2.5E-02 Thyroid 4.2E-01 1984 Update 4
1 Table 15.6.5-1A , 1 Parameters f or Post-LOCA Suberit icalit y Analysis, ; Volume Groupina Boron concentration (ppm) Low Head Safety Injection-(LHSI) Discharge ! to Interraediate Head Safety Injection (IHSI) - and High Head Safety Injection (HHSI) suction (Valve nil 36B to Valves NI332A & NI333B) 1980 Refueling Water Storage Tank (RWST) to Valve FW28 FWST to IHSI suction r FWST to Valve Hv223 Normal Containment Spray Discharge -! Containment Spray Suction from RWST LHSI Discharge to Aux. Cont. Spray i LHSI Suction from Sump 350 2 i LHSI Suction from Loop C Hot Leg ! Containment Spray Suction from Sump RCS LHSI Discharge to Cold Legs LHSI Discharge to IHSI and HHSI Suction (Valve ND58 to Valves NI332A & NI333B) , (LHSI Discharge to Valves ND58 & nil 36B) LHSI Discharge to B and C Hot Legs Valve FW28 to LHSI Suction variable' LHSI Mini-Flow IHSI Discharge to LHSI Discharge IHSI Discharge to Hot Legs IHSI Mini-Flow HHSI Discharge to Cold Legs
- Valve NV223 to HHSI Suction I
t l i
' EOC Mode 4 RCS boron concentration
" variable" indicates that the-associated volume concentration is assumed equal to the RCS boron concentration, which is a functioni of burnup.
1 n-,4a ,- a , _.--,,e, ,, .-cn-,_.,,n ..,----.e,,,,,.,.ww-,-,-.-e., , ,,,-,~mn,,n __.,,g>, ,,n,,,,
- - , , -,rry :q r v.< me,,--w,, - - - - +
5 1 Table 15.6./-9 (Page 1 of 3) Parameters for LOCA Of fsite Dose Analysis
- 1. Data and assumptions used to estimate radioactive source from postulated accidents Dill
- a. Power Level (MWt) M
- b. Failed fuel 100% of fuel roos in core
- c. , Activity released to containment atmo-sphere from failed fuel and available for release (percent of core activity)
Noble gases 100 Iodines 25
- d. lodine activity released to containment 50 sump and available for release via ECCS leakage outside contaiment (percent of core activity)
- e. Iodine fractions (organic, elemental, Regulatory Guice 1.4 and particulatel
- 2. Data and assumptions used to estimate act-ivity released
- a. Containment free volume Upper containment volume (ft2) 6.70E+05 Lower containment volume (f t2) 3.68E+05 Total
- containment free volume (ft2 ) 1.03SE+06
- b. Containment leak rate (percent of containment volume cer day) 0< t < 24 hrs 0.3
-}}