ML20209H033
| ML20209H033 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 07/13/1999 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20209H025 | List: |
| References | |
| NUDOCS 9907200104 | |
| Download: ML20209H033 (49) | |
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*6- " M dF Recirculation Loops Operating 3.4.1 t
) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating b LCO 3.4.1 The reactor core shall not exhibit core thermal hydraulic instability or operate in the " Scram" or " Exit" Regions.
"I 4
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> a. Two recirculation loops with matched recirculation loop jet pump flows shall be in operation:
M {
- b. One recirculation loop may be in operation provided:
- 1. LC0 3.3.1.1 " Reactor Protection System (RPS)
Instrumentation." Function 2.b (Average Power Range Vl Monitors Simulated Thermal Power-Upscale) Allowable Value of Table 3.3.1.11 is reset for single loop e operation, when in MODE 1: and k 2. THERMAL POWER is s 67.2% RTP.
.J % q ................~............N0TE - --- -- - - ---- ---
Required allowatt.e value modification for single loop x j El operation and THERMAL POWER limitation may be delayed for up I L to 4 hours after transition from two recirculation loop operations to single recirculation loop operation. APPLICABILITY: MODES I and 2.- ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation jet pump A.1 Declare recirculation 2 hours 100) flow mismatch not loop with lower flow: wit 11n limits. "not in operation." (continued) 9907200104 990713 PDR , P ADOCK 05000341 i pg l FERMI - UNIT 2 3.4 1 Revision 10. 07/09/99 l
Recirculation Loops Operating B 3.4.1
) BASES APPLICABLE SAFETY ANALYSES (continued)
Thermal hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal hydraulic instability are necessary to provide reasonable assurance that the requirements of Reference 4 are satisfied. Recirculation loops operating satisfies Criterion 2 of 10 CFR 50,36(c)(2)(ii). LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.2 to ensure that during a LOCA caused by a break of m
~
the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in i SR 3.4.1.2 not met, the recirculation loop with the lower d flow must be considered not in o With only one
% recirculation loop in operation,peration.
modifications to the APRM Simulated Thermal Power-Upscale setpoint (LC0 3.3.1.1) and y{ a limitation on THERMAL POWER may be applied to allow S continued operation consistent with the assumptions of the safety analysis, f e Operations that exhibit core thermal hydraulic instability ) are not permitted. Additionally, in order to avoid G potential power oscillations due to thermal hydraulic instability, operation at certain combinations of power and
?u flow are not permitted. These restricted power and flow cc regions are referred to as the " Scram" and " Exit" regions y and are defined by Bases Figure B 3.4.1-1.
A Note is provided to allow 4 hours following the transition to single loop operation from two loop operation to jl establish the APRM Simulated Thermal Power - Upscale m setpoint in accordance with the single loop allowable value. ' dt which is specified in Table 3.3.1.1-1 and to establish 9C operation at 5 67.2% RTP. The 4 hour period is sufficient to make the adjustments given the relatively small change j required. This transition that results in a> plying the new 4 single-loo) allowable values to APRM OPERABI_ITY, is such ! g that any ARPM non compliance with the required allowable I
> value after.this 4 hour allowance results in ACTIONS of ;
4 LC0 3.3.1.1 being entered: no ACTION of LC0 3.4.1 would l apply. j -- l l FERMI - UNIT 2 B 3.4.1 - 4 Revision 10. 07/09/99 I l I
$ fEC IFt cAgian) 3 <{. /
3 /4. 4 REACTOR COOLANT SY' STEM [Nse Su SFi eq/fori 54/0) 3 3/4.4.1- RECIRCULATION SYSTEM 1 RECIRCULATION LOOPS < LIMITINGtCONDITION FOR OPERATION l.M p ,g 2.t.:.1- Two reactor coolant system recirculation loops shall be in operation. APPLICABILITY: OPERATIONAL CONDITIONS 1-and 2*. ACTION:
- a. With one reactor coolant system recirculation loop not in operation:
LLo M .I' 3. Within 4 hours: LAil t@TW~ - ra m)Plar/ the individual op#atino rarAculation recirculatio(pumo pump in the Manual flow cof roller'for the ] g moda g y gy,g,Qb) Reduce THERMAL POWER to less than or equal to 67.27, of RATED THERMAL POWER. j ,j
) Limit the speedff the operating recire or'eaual to 751f of rated pump speeds)/Intion pump to les/than]
^
d) in ase Ine MiniriurycMITICAL Powtx MAisy (Isrn) barety L1 t to] value for sing)e loop operation reqVired by Specifica on j
.27 g
3 e) Change the Average Power Range Monitor (APRM) Simulated Thermal, n j
@ N'I'g'g Power - Upscale Flow Biased Scram tr.f N :".en Nr se- rMis?
WA110wable Values to those applicable for single recirculation ~ loop operation per Specifications 2.2.1 and 3.3.6. l f '"ferform Surveille.nce Requirement 4.4.1.1.4 if THERMAL POWER is gggfugh7 )f recirculation loop flow in the operating loop is le gg ,,, less than or equal to 307, of RATED THERMAL POWER or the i u equal to 507, of rated loop flow,
-2. Oth:r "-- 6 4- it h ::: ll0T ;;rJ ;; n 3
. 6uin tue u m n ,,_ . . .
- b. With no reactor coolant system recirculation loop in operation while in bTION D OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.
- c. ~ With'no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least
$ttTtotJC. HOT SHUTDOWN within the next 6 hours. ,
{See cialTe3rException3.f0.4.] l e -' FERMI - UNIT 2 3/4 4-1 Amendment No. JJ,64,EJ,EJ, E7.JES, 122 PAGE 2. OF 06 ggyto i
SPEttFicATigd 3.% I (Ais.sa:rpafwrw 3 Go) L (Mso sa sp;(;w% s E d l REACTOR COOLANT SYSTFM
)
l SURVETLIANCE REQUIREMENTS S e4. T4.1.1.1 Each pump discharge valve shall be demonstrated OPHRABLE by j
$bbicycling each valve through at least one complete cycle of full travel at least J 3.g,i once per 18 months.
l
-4 A .I.1.2 LEO r.u / h, l d.4.1.1.3 With ne reactor coolant s tem recmuieuon ioop not in
'q eration, at I st once per 12 hours erify that: '
. THERMAL POWER i ess than or equal to 6 .2% of RATED )
THERMAL POWER, nd
- b. The individu recirculation pump flo controller for t e 1 operating r irculation pump is in e Manual mode, a j
- c. The speed f the operating recircu tion pump is le' than or coual o 75% of rated oumo see -
b4.4.1.1.4 With one reactor coolant system loop not in operatjon with THERMAL 1 POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of r4ted loop flow, verify the following differential temperature requirements are met within no 1 more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase:
) M
- a. Less than or equal to 145'F between reactor vessel steam space coolant and bottom head drain line coolant, and ged[(cqh'A b. Less than or equal to 50'F between the reactor coolant within the loop not in operation and the coolant in the 3 Y.rb / reactor pressure vessei *, and
- c. Less than or equal to 50*F between the reactbr coolant within the loop not in operation and the operating loop.**
b*$ *y. m ** Requirement does not apply when the recirculation loop not ih operation is 1 5p46u.h> isolated from the reactor pressure vessel. 3M6 FERMI - UNIT 2 3/4 4-2 Amendment No. 83. 59. 87. US.133 PAGE 3 _OF 06 ggqlo
r DISCUSSION OF. CHANGES ITS: KECTION 3.4.1 - RECIRCULATION LOOPS OPERATING TECHNICAL CHANGES - NORE RESTRICTIVE Non. TECHNICAL CHANGES'- LESS RESTRICTIVE
" Generic" LA.1 CTS 3.4.1.1,~ Actions a.1.a) and c), and 4.4.1.1.3, impose limitations on: the recirculation pump flow controller mode of operation; and the operating recirculation pump speed. These limitations are g' related to operational considerations associated with
. prevention of possible control oscillations and reactor I vessel internals vibration, and are not associated with the function of the Recirculation Loop Operating ~
Technical Specification as defined in the NUREG-1433 4 Bases. Therefore, the requirements for maintaining M these limitations can be adequately defined and - controlled in the Technical Requirements Nanual (TRM), j[ which require revisions to be controlled by 10 CFR 50.59. These relocations continue to provide adequate > protection of the public health and safety since the ITS retain sufficient requirements related to maintaining thermal limits and thermal hydraulic stability for single. loop operation. LA.2 Not used i I O FERNI - UNIT E2 3 REVISION 10, 07/09/99
Recirculation Loops Operating 3.4.1
') 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating d@b k LCO 3.4.2 recirculatio oops with matched flows sha i be in M'8' operation, rew c,geg( n + g;),;p 3*OI 3) core % mot h'idasuc iris la,Wllfy oofetuk- -
Y in Se " Scrod or %ii% gin
>0ne recirculation loop may be in op pro
, , 4mpts-we-applied-when-the-essoci et
~ 0:.:.:,.= =.~ = - y m = =-
( HGR)," single loo operation limits pecified in F
.. 4 eCOLRp l W
, LCO 3.2.2, " MINI CRITICAL POWER 10 (MCPR)," si gl o operation mits Isoecifiaa in +k. entp1 and av n
CO 3.3.1.1, " Reactor Protection System (RPS) u MC f Instrumeritation." Function 2.b (Average Power an e / Achm MonitorslFle hudhimulated Thermal Power Allowable Value of Tabl_M1.1-1_is reset for s ngle
, N aa.e)\ f loop operatio ) ( ppe , /> ?
APPLICABILITY: V' lME(LT MODES I and 2. L 3 41- l j' ?e ACTIONS @ & CONDITION REQUIRED ACTION COMPLETION TIME Reg rement of the A.1 Satisfy t 24 hou LC not a . require nts of the LCO.
@ /
( --nm. .; ....:: .;, :=. (continued)
/tdsEET 3.y.1- 2.
BWR/4 STS 3.4-1 Rev 1, 04/07/95 w4 " t
I Recirculation Loops Operating 3.4.1 INSERT 3.4.1 - 1 LC0 g 3.4.1.1 J b. 2. THERMAL POWER is s 67.2% RTP.
..........................N0TE-----
Required allowable value modification ------------------ for single loop -- g 3> 4 operation may be delayed for up to 4 hours after transition from two recirculation loop operation to single y.q,t.1 recirculation loop operation. Ac. nog a.1 FERMI - UNIT 2 Page 3.4 1 (1) (INSERT) REVISION 10, 07/09/99l
Recirculation Loops Operating B 3.4.1
.. BASES .s I
APPLICABLE A plant specific LOCA analysis has been performed assuming SAFETY ANALYSES only one operating recirculation loop. This analysis has (continued) demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core coo { . j ==; revi,: le er. ::fified The transient analyses of Chapter 15 of the SAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed,;r n !d;d 1; 7.^7; P3 r:; i - -te a-a d 'f-d. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the diffe t , relationships between recirculation drive flow and . ector
'2, . , The A" E P. trd ". = eetveir.t; fr :te:h _!ee n
'- fI core flow.:;: :t!?
r
?- !!? !ed fr t.' % ;. The APRM W W fimulated T)(FJD%I tsetpoint is in LCO 3.3.1.1, " Reactor Inferi Protection 5fsteli ( 5 Instr 5==a+=tlan."
g3 3.v,/-2. I- >Recirculatlan loops - O[5 c* lC.
-- M t
=^ Fol k, i.i ..t.
to cfR sct3HeXaf]erion 2 of 4he-3 i LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1A to ensure that during a LOCA caused by a break of i sne piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in 5R 3.4.I V not met, the recirculation loop with the lower flow must be considered not in operation. With only one I'Cfleu}agg,n,,1, opp gogyation ,mogtiop,to the,,__ k' .. ... .. ...... - . . . . . m - - . u ,,um HEAHrENERAHC% MTC (',PZP.)'), "= 10::it: (' 00 3.2 ' .
' [ehnt"(50 applied to allow continuea operation consistent with the 31 assumptio of Inse rt p,4 a e:.::3.
^ ^ ^
o sid only y v-& ondUM{aN on W EtM N f 4 % 4 6 3 41-3 } G (continued)
-" j: liY3- B 3.4-3 ,ev 1, ^4/"Jiii s
Ra/10
r 1 Recirculation Loops Operating B 3.4.1 m l l Insert B 3.4.1 - 2 1 Thermal-hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal-hydraulic instability are necessary to provide reasonable assurance that the requirements of Reference 4 are satisfied. 1 1 Insert B 3.4.1 -3 Operations that exhibit core thermal-hydraulic instability are not permitted. Additionally, in order to avoid potential O power oscillations due to thermal-hydraulic instability, l operation at certain combinations of power and flow are not permitted. These restricted power and flow regions are referred to as the " Scram" and " Exit" Regions, and are defined f. s by Bases Figure B 3.4.1-1. y l l A Note is provided to allow 4 hours following the transition to single loop operation from two loop operation to establish k T the APRM Simulated Thermal Power-Upscale setpoint in y accordance with the single loop allowable value, which is specified in Table 3.3.1.1-1 and to establish operation at s; 67.2% RTP. The 4 hour period is sufficient to make the \y adjustments given the relatively small change required. This transition that results in applying the new single-loop D l allowable values to APRM OPERABILI7, is such that any APRM i non compliance with the required aFlowable value after this f 4 hour allowance results in ACTIONS of LC0 3.3.1.1 being entered; no ACTION of LC0 3.4.1 would apply. l$e 9 l l FERMI UNIT 2 Page B 3.4 3 (INSERT) REVISION 10. 07/09/99l l
H . SRVs B 3.4.3
) BASES p ;
- APPLICABLE The~ overpressure protection system must accommodate the most
' SAFETY ANALYSES- severe pressurization transient. Evaluations have determined that the most severe transient is-the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref.1). For the purpose of the analyses 11 SRVs are assumed to operate in
.the safety mode. The analysis results demonstrate.that the design SRV capacity is capable of maintaining reactor pressure below the ASE Code limit of 110% of vessel design pressure 1(110% x 1250 psig
- 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during.
the Design Basis Event. From an over>ressure standpoint, the design basis events are bounded by t1e MSIV closure with flux scram event described above. Reference 2 discusses additional events that are expected.to actuate the SRVs. SRVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). I LC0 The safety function of 11 SRVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. I and 2). The. requirements of this LCO are applicable only to the capability of the SRVs to mechanically open to relieve - excess pressure when the lift setpoint is exceeded (safety function). The SRV setpoints, and 3% allowance for setpoint drift. are Ok~l , established to ensure that the ASME Code limit on peak reactor pressure.is satisfied. The ASME Code specifications require the lowest safety valve set vessel design pressure (1250 psig) andpoint to be atsafety-the highest or below valve-to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the UFSAR are also based'on these setpoints. Operation with fewer valves OPERABLE than specified, or with i setpoints outside the ASME limits, could result in a more i severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor
... pressure being exceeded, i
?
l FERMI:- UNIT 2 - B 3.4.3 -2 Revision 10 07/09/99
S/RVs B 3.4.3 f BASES APPLICABLE pressure (110% x 1250 psig - 1375 psig). This LC0 helps to SAFETY AMALYSES ensure that the acceptance limit of 1375 psig is met during (continued) the Design Basis Event. From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses additional events that are expected to actuate the S/RVs. Op,f S/RVs satisfy Criterioni_b 3- of 'M 50.$ t^
""! "eli-{--
CTR _- Ql'2 Ote% D )- c LC0 The safety functiori ofX1lyS/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. I and 2). The requirements of this LC0 are applicable only to the capability of the S mechanically open to relieve excess pressu/RVs re whentothe lift setpoint is exceeded,Jsafety faartioni e - Op*q aw3 3) TheS/RVsetpoints.(.r,..staolisneasoensurethatthee AttoGwet. Or algaid d Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for ove ssurization conditions. The transie_n valuations in
] - .
t arW+basedonthesesetpoints;4 sty"ofeehade-tta
- f ' f-9 ree ':fet! : a' + " a' na -
2:. ;;tp'.;.t l dei't tr ;n;f t :: :'f:f t;n; ;f ;;;.;;r;;ti;;;. Operation with fewer valves OPERABLE than specified, or with ; setpoints outside the ASME limits, could result in a more ' severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded. 1
- l
,l APPLICABILITY U
InMODES1,2,and3.hS/RVsmustbeOPERABLE,since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of i dissipating the core heat. (continued) BWR/4-STS- B 3.4-13 Re; 1,- 0"j^7/n-
RCS Operational LEAKAGE 3.4.4
) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LC0 3.4.4- RCS operational LEAXAGE shall be limited to:
- a. No pressure boundary LEAKAGE:
- b. 5 5 gpm unidentified LEAKAGE:
- c. s 25 gpm total LEAKAGE averaged over the previous 24 hour period: and
- d. s 2 gpm increase in unider . ,fied LEAKAGE within the previous 24 hour period in MODE 1.
APPLICABILITY: MODES 1, 2. and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME s A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours not within limit, within limits. M Total LEAKAGE not within limit. h B. Unidentified LEAKAGE B.1 Reduce LEAKAGE to 4 hours t increase not within within limits.
? limit.
k 2 (continued) l l l- FERMI - UNIT 2 3.4 8 Revision 10, 07/09/99 j
l
, . l RCS Operational LEAKAGE
$ 3.4.4 l
) ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME j g i B. (continued) B.2 Verify source of 4 hours unidentified LEAKAGE increase is not 4 service sensitive type 304 or type 316 austenitic stainless steel.
C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A 6tLD > or B not met. I C.2 Be in MODE 4. 36 hours 2 Pressure boundary LEAKAGE exists. . i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY fl SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increase are 8 hours within limits. l FERMI UNIT 2 3.4 9 Revision 10, 07/09/99
r (c
- L RCS Operational LEAKAGE B 3.4.4
) BASES APPLICABILITY In MODES 1, 2. .and 3 the RCS operational LEAKAGE LC0 a> plies, because the potential for RCPB LEAKAGE is greatest w1en the reactor is pressurized.
In MODES 4 and 5. RCS operational LEAKAGE limits are not required since the reactor-is not pressurized and stresses l .in the RCPB materials and potential for LEAKAGE are reduced. R -ACTIONS &J l With RCS unidentified or total LEAKAGE greater than the
. limits, actions must be taken to reduce the leak. Because the LEAKAGE limits are conservatively belos the LEAKAGE that would constitute a critical crack size, 4 hours is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as h
identified LEAKAGE:.however, the total LEAKAGE limit would remain tuhanged.
' kl - B.1 and B.2 An unidentified LEAKAGE increase of > 2 gpm within a 24 hour-t period is an indication of a potential flaw in the RCPB and
!- must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to be the source of the LEAKAGE increase within the required Completion Time by evaluating service sensitive type 304 and
. type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or
- intermittent flow fluids and determine it is not the source I of the increased LEAKAGE. This type piping is very I
susceptible to IGSCC. For an unidentified LEAKAGE increase l greater than required limits (in accordance with LCO 3.0.2), an alternative to this evaluation is to reduce the LEAKAGE-increase to within limits (i.e.. reducing the LEAKAGE rate such that the current rate is less than the "2 gpm increase i in the previous 24 hours" limit: either by isolating the l source or other possible methods). The 4 hour Completion Time is reasonable to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety. l FERMI . UNIT 2 - B 3.4.4-4 Revision 10, 07/09/99 l
RCS Operational LEAKAGE B 3.4.4
} BASES ACTIONS (continued)
C.1 and C.2 l If any Required Action and associated Completion Time of Condition A or B is not met or if pressure ~ boundary LEAKAGE exists, the plant must be brought to a MODE in which the LC0 does not apaly. To achieve this status, the plant must be brought to 10DE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner , and without challenging plant safety systems. SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE (e.g., Primary Containment Atmospheric Gaseous Radioactivity, RPV head flange leak detection, and sump monitoring systems). Leakage detection instrumentation is discussed in more detail in the Bases for LC0 3.4.6 "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates: however, any method may be used to quantify LEAKAGE within the guidelines 3 of Reference 5. In conjunction with alarms and other Tl administrative controls, an 8 hour Frequency for this Surveillance is appropriate for identifying LEAKAGE and fur kl tracking required trends (Ref. 6). REFERENCES 1. 10 CFR 50, Appendix A. GDC 30.
- 2. GEAP 5620, April 1968.
- 3. NUREG-76/067, October 1975.
l
- 4. UFSAR, Section 5.2.7.4.3.3.
- 5. Regulatory Guide 1.45.
- 6. G0neric Letter 88-01, Supplement 1.
l FERMI - UNIT 2 B 3.4.4-5 Revision 10, 07/09/99 i
Sfec4cotwa 344 D
.p REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued)
- e. In OPERATIONAL CONDITION 1, with any reactor coolant system
[ g , art $ UNIDENTIFIED LEAKAGE increase greater than 2 gpa within any 24
' hour period, identify the source of leakage increase as not service sensitive Type'304 or 316 austenttic stainless steel
- within 4 haursfor be in at least HOT SHUTDOWN within the next 12 Action C hours and in COLD SHUTDOWN within the following 24 hours.
- f. In OPE L COND 2 and (with any ctor coo nt system v L.2 UNID FIED L increase ater tha gpm wit any 4 ur I p d, ident he sour f leakage rease et serv nsitive 304 or austeniti ainles eel wi 4 /
hours in at - t HOT SHUT withi e next in SHUTDOW thin the fo wing 24 ours j- hoursand} SURVEILLANCE REOUIREMENTS
~3 4 -} -4.4.2.2.b The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
o e ne d en d =n+ We,qese'o*'u LA.I , dio vit[6tleastonceperenours1_ bgo#DnaAne primary corttetsment-tump -tro er hours th vrt.MilDML wnplIIUM ItVand at least once per 4 hour (iin OPERATIONAL CONDITIONS 2 and 3, l d 3 _ T
/ [c #enftorgthe.ArfGeMacMn4Mel hourst*rin OPERA"IONAL CONDITION W and at least#once 1dOper 4 [j 'f
< i ours in OP7 RATIONAL CONDITIONS 2 and 3, and ,,
rin AdakAMM% LA. ) l I lad l [m wrai .*.rsh1'T7vdM
. . a . m. . g m m,. 4
.s M " - - - r.nn u a-..t
. c. ; . ,.
in emn n=: - r -- = ; . m a u l L,j t I <- FERMI - UNIT 2 3/4 4-11 Amendment No. 89 1
%m' PAGE- c2 0F 86VlO 02
DISCUSSION OF CHANGES ITS8 SECTION 3.4.4 - RCS OPERATIONAL LEAKAGE TECHNICAL CHANGES - LESS RESTRICTIVE O
" Specific" 2.
L.1 CTS 4.4.3.2.1 rquires that leakage be demonstrated to be within limit,in N part, by monitoring primary containment atmospheric gaseous radioactivity at least once per 4 hours. However, the remaining parts of CTS 4.3.2.1 l require leakage be demonstrated to be within limit by monitoring once per 12 hours; and this 12 hour monitoring is done on the system that actually quantifies the leakage (the atmospheric gaseous radioactivity monitor is p_ot o utilized to quantify the leakage for comparison to the LCO limit, as provided in CTS footnote
- to the referenced surveillance). Note also that in Mode 1 l this Frequency is restricted from applying the 25% extension of ITS SR 3.0.2.
ITS SR 3.4.4.1 requires verification every 8 hours that the RCS unidentified and total Leakage, and unidentified Leakage increase, are within limits (which is more restrictive than these latter CTS requirements, but included here for completeness). RCS Leakage is monitored by a variety of instruments designed to provide alarms when excessive Leakage is indicated and to quantify the various types of Leakage. In conjunction with alarms and administrative controls, an 8 hour Frequency for this Surveillance is appropriate for identifying Leakage and for tracking trends. With this increased Frequency of performance (from 12 to 8 hours), the 25% extension of SR 3.0.2 is allowed to be applied (i.e., a 10 hour maximum interval; still
. more restrictive than the CTS 12 hours). This change is consistent with the intent of Generic Letter 88 01, to provide an effective means to determine any adverse trends and as such will have a negligible impact on safety.
Q
;3 FERMI - UNIT 2 2 REVISION 10, 07/09/99
. RCS Operational LEAKAGE 3.4.4 3 3.4 REACTOR COOLANT SYSTEM (RCS)
I 3.4.4 RCS Operational LEAKAGE CIS} LCO 3.4.4 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE;
- b. s gpa unidentified LEAKAGE 3**
p*l c. s fSS) gpm total LEAKAGE averaged over the previous 24hourperiod;fand
- d. s 2 gpa increase in unidentified LEAKAGE within the l previous M hour period in MODE 1.K j APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours not within limit, within limits. ,3,7 Achto!L b. Total LEAKAGE not within limit. B. Unidentified LEAKAGE increase not within B.1 Reduce LEAKAGE to within limits. 4 hours g limit.
\
.Qll N t
(continued) 1
- "":n/4 373 -
. 3.4-7 En 1, G4/G7/55 e#
1
l RCS Operational LEAKAGE 3.4.4 ACTIONS / CTS) CONDITION REQUIRED ACTION COMPLETION TIME y i B. (continued) B.2 Verify source of 4 hours unidentified LEAKAGE 3.4. 3. 2 increase is not , service sensitive dc,htoftd. type 304 or type 316 austenitic stainless steel. C. Required Action and C.1 Be in MODE 3. associated Completion 12 hours Time of Condition A AEl 3.4.3.1 or B not met. d#^ #> C.2 Be in MODE 4. 36 hours b /6 DE Pressure boundary LEAKAGE exists. .c SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE S hours and unidentified LEAKAGE increase are ,4.3,2.tk G within limits. / g 3L"'/t :;T3 3.4-B -Rev 1, O'/07/o t ./ Rev' W 4
RCS Operational LEAKAGE B 3.4.4 f- BASES (continued) ACTIONS L1 With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leak. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE; however, the total LEAKAGE limit would remain unchanged. y i k B.1 and B.2 An unidentified LEAKAGE increase of > 2 gpa within our period is an indication of a potential flaw in the RCPB and 1 must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to _ the source of the LEAKAGE increase within the required un accorda m Completion Timer)for an unidentified LEAKAGE increase w;w r o.z ') greater than required limitsf an alternative to ' '-- ^ LEAKAGE increase to within limits (i.e., reducing the thi5e*Lution zy LEAKAGE rate such that the current rate is less than the Qsto %ct fhe, "2 com increase in the previou hours" limit; either by ~ Op'q isolating the source or other p ss ble methods? is::ta=. p) r-+m service sensitive type 3v, anc type 3e6 austenttici est{afj stainless steel iping that is subject to high stress or that contains re atively stagnant or intermittent flow fluids and determine it is not the source of the increased LEAKAGE. This type piping is very susceptible to ISSCC. The 4 hour Completion Time is reasonable to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety. C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within It hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, (continued) nun /A NS-- B 3.4-20 Rr; I, 0i/07/95 Rev10
I RCS Operational LEAKAGE B 3.4.4 5- BASES
]
ACTIONS C.1 and C.2 (continued) based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems.
=
les Con kinme.<tMmos JAra 6esem.s SURVEILLANCE SR 3.4.4.1 /d.oum(/ae f,V;/yl,, R,wbv Ata,I 8/ age lee.le c(4 1 REQUIREMENTS men i e e M _s vs1eis.3 - The RCS LEAKAG s mon ored by a vari y of instrumiints - designed to provide alarms when L is indicated and to quantify the various types of LEAKAG Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.6, 'RCS Leakage Detection Instrumentation." Sump
' level and flow rate are typically monitored to determine actual LEAKAGE rates; however, an quantify LEAKAGE within the guide'y Jines method may 5.
of Reference be used In to .
' conjunction with alarus and other administrative controls, an 8. hour Frequency for this Surveillance is appropriate for identifying LEAKAGE and for tracking required trends [
(Ref. 6). I REFERENCES 1. 10 CFR 50, Appendix A. GDC 30. r- )
./ 2. GEAP-5620, April 1968.
- 3. NUREG-76/067, October 1975.
,2 .Y.
.4. yFSAR, Section .".. ...
- 5. Regulatory Guide 1.45.
- 6. Generic Letter 88-01, Supplement 1.
. n.,, _
?""f' !!! -
B 3.4-21 Rsi ;, O'/0:'/0 fievl0
JUSTIFICATION FOR DIFFERENCES FROM NUREG - 1433 ITS: SECTION 3.4.4 - RCS OPERATIONAL LEAKAGE NON BRACKETED PLANT SPECIFIC CBANGES P.1 These changes are made to NUREG 1433 to reflect Fermi 2 current l licensing basis: including design features, existing license requirements and commitments. Additional rewording, reformatting, and revised numbering is made to incorporate these changes consistent with Writer's Guide conventions. ! P.2 Bases changes are made to reflect plant specific design details, I equipment terminology, and analyses. P.3 Not used. P.4 Change made for editorial preference or clarity. h P.5 Not used. d P.6 The reference to the NRC Policy Statement has been replaced with a lh more appropriate reference to the Improved Technical Specification
" split" criteria found in 10 CFR 50.36(c)(2)(ii).
)
4 FERMI UNIT 2 1 REVISION 10, 07/09/99l
1 I RCS Leakage Detection Instrumentation 3.4.6
] 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Leakage Detetion Instrumentation LC0 3.4.6 The following RCS leakage detection instrumentation. shall be OPERABLE:
- a. Drywell floor drain sump flow monitoring system:
d b. The primary containment atmosphere gaseous radioactivity 1 monitoring system channel; and
- c. Drywell floor drain sump level monitoring system.
APPLICABILITY: MODES 1, 2, and 3. ACTIONS
.................................. NOTE- ---- ---- - - ------------- ---------
LC0 3.0.4 is not applicable. s CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell floor drain A.1 Restore drywell floor 30 days sump flow monitoring drain sump flow system inoperable. monitoring system to OPERABLE status.
.B. Required primary B.1 Analyze grab samples Once per containment atmosphere of primary 24 hours gaseous radioactivity containment monitoring system atmosphere.
inoperable. (continued) j l FERMI UNIT 2 3.4-13 Revision 10, 07/09/99
1 RCS Leakage Detection Instrumentation ! 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Leakage Detection Instrumentation [ cts) l d LCO 3.4.6 The following RCS leakage detection instrumentation shall be I OPERABLE. [
- a. Drywell floor drain sumpfmo to ing system:M(7.4,J.l.b h
- b. ^ - ""---' - pr imary containment 6
*g cat-r em+-matmosphe gaseousmonitoring_systeg;f.43I.
c. Stand a 9foyg@g,,g noe r cooler 3pertiensajed1pate] h DrfN *LE Noor drain surp APPLICABILITY: MODES 1, 2, and 3. '###$ "'*0 N##'"j 5[8b- g ACTIONS f 3 CONDITION
\ REQUIRED ACTION COMPLETION TIME
\
A. Drywell floor drain ----
-NOTE-------- D g sumpamonitoring system LCO 3.0.4 is not applicable. 3,4,3 l !
I j inoperable. . . Ac-l.io tt A.I Restore drywell floor 30 days drain summ monitoring system tof0PERABLE status. (continued)
.-0Wtf* T.r- 3.4-12 En 1, 54/07/fr5-
~~' W fD
RCS P/T Limits 3.4.10
] SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.10.5 -- ------- --- - - NOTE-- ------ ---------- Only required to be met during a THERMAL POWER increase or recirculation flow
/ inc.' ease in MODES 1 and 2 with one idle recirculation loop when THERMAL POWER is s 30% RTP or when operating loop flow is s 50% rated loop flow.
L Once within Verify the difference between the bottom 15 minutes l head coolant temperature and the RPV steam prior to a space coolant temperature is s 145'F. THERMAL POWER l increase or recirculation l l flow increase SR 3.4.10.6 --------- --- ---- NOTE- --- - - ------- - Only required to be met during a THERMAL POWER increase or recirculation flow i increase in MODES 1 and 2 with one non-isolated idle recirculation loo) when THERMAL POWER is s 30% RTP or w1en l operating loop flow is s 50t rated loop flow.
........................................... Once within
. 15 minutes 4 l Verify the difference between the reactor prior to a coolant temperature in the idle THERMAL POWER
/ . recirculation loop and the RPV coolant increase or temperature is s 50*F. recirculation 3f l flow increase l
l (continued) ! 1 I I l FERMI.- UNIT 2 3.4 26 Revision 10 07/09/99
o ,_ . r E RCS P/T Limits B 3.4.10 i l } ' BASES SURVEILLANCE. REQUIREMENTS (continued) , This SR has been modified with a Note that requires this l- Surveillance to be performed as applicable only during J system heatup and cooldown operations and inservice leakage and hydrostatic testing. < SR 3.4.10.2 l A separate limit is used when the reactor is approaching ) I criticality. Conse uently, the RCS pressure and temperature must be verified wi hin the appropriate limits before withdrawing control rods that will make the reactor critical. Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality ; provides adequate assurance that the limits will not be j exceeded between the time of the Surveillance and the time ! of the control rod withdrawal. l SR 3.4.10.3. SR 3.4.10.4. SR 3.4.10.5. and SR 3.4.10.6 3 Differential temperatures within the applicable limits l
~j' ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
Limiting _ differential temperatures within the applicable limits during a THERMAL POWER increase or recirculation flow
-increase in single loop operation, while THERMAL POWER s 30t RTP or operating loop flow s 50% of rated loop flow, ensure i that thermal stre b es resulting from THERMAL POWER increases or recirculation loop flow increases will not exceed design !
- allowances.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump. THERMAL POWER increase i during single loop operation, or recirculation flow increase during single loop operation, provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start, power increase. or flow increase. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 and SR 3.4.10.6 is to compare the temperatures of the operating
-recirculation loop and the idle loop.
J .. l FERMI - UNIT 2 B 3.4.10 - 7 Revision 10, 07/09/99
S fE c t r-t c N n M 8 .'h l O (MSOSeeSecTScaHM
/ 3 4. /)
lse? Set Geci $~CaNM 8 0~'I) REACTOR COOLANT SYSTEM SURVETLLANCE RFOUIREMENTS
- 4 (8 b cy.4.1.1.1 Each pump discharge valve shall be demonstrated OPGRABLE by Si cling each valve through at least one complete cycle of full travel at least 3Sil once per 18 months. '
i.4.1.1.2 DELETED 4.4.1.1.3 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours verify that: ' Sq a. THERMAL POWER is less than or equal to 67.2% of RATED THERMAL POWER, and
%cifuah% b. The individual recirculation pump flow controller for the 3.Y. I .
operating recirculation pump is in the Manual mode, and
- c. The speed of the operating recirculation pump is less than or equal to 75% of rated pump speed.
4.4.1.1.4 With one reactor coolant system loop not in operation with TilERMAL POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of r$ted loop flow, verify the following differential temperature requirements are met within no more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase: GC3. 4 lD.I a. Lets than or equal to 145*F between reactor vessel steam I space coolant and bottom head drain line coolant, and
- b. Less than or equal to 50'F between the reactor coolant M M.lo .h wi_tttin the loop not in operation and (the coolant in th
. Qeactor pressure vesset=. ano A
- c. Less than or equal to 50'F between the reactbr/ coolant within the loop not in operation andine operating loop)**
4
--*"at u seh
** Requirement does not apply when the recirculation loop not ih operation is 3.'f.la.h isolated from the reactor pressure vessel.
~ FFRMI - UNIT 2 3/4 4-2 Amendment No. 53. 59. E7. UB,133 PAGE c1 0F 08 RW 0
r I RCS P/T Limits 3.4.10 SURVEILLANCE REQUIREMENTS (continued) M l i SURVEILLANCE FREQUENCY l C'I SR 3.4.10.3 - --- NOTE- ----- Only requ.. uired
.no 4,p .. .to.be
- . . met
. . .-inaMODES
_ .. . . ..1, 2, 3 h ' YO; gig) l l Auti (tLitcal 'on - ,~ er ,.i.,X,. " ~~~~~ ~~ ~ " ~~~~~~- L tu y s M qa Verify the difference between the bottom Once within y , head coolant temperature and the reactor 15 minutes h.95)M SPact, pressure vessel (RPV)fcoolant temperature prior to each is Mth .: li:::it: :pe:!'!ed 4'
+he DTLD startup of a (iqg,y recirculation pump SR 3.4.10.4 ---------------NOT E------------- -
Only required to be met in MODES 1, 2, 3, [3.4l#, AglicW/f/y) dvrG, it - - - - - - - - - --- - - - - - - - - ppp forkup, -- - 3. 4 \ d Verify the difference between the reactor Once within 4.4. l . M coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature
-in the PTLR. yi 1-t (Mke 14-it: spt:i't:d-f5oy startup of a recirculation lt5 E W B. % to - 2 #
SR 3.4.10 ------------------NOTE-------------------- I Only required to be performed when tensioning the reactor vessel head bolting " 4 fal.'l b ) studs. Verify reactor vessel flange and head 30 minutes flange temperatures are ti.hin +he 'i=i+3 J h PTL".J> il s p When
&::i't:dinth:kc4x vedeIhead 5H dvd.s are [ j (continued)
BWR/4 STS 3.4-25 Rev 1, 04/07/95 RW LO
I RCS P/T Limits 3.4.10 i
~
INSERT 3.4.10 _2 j l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.5' ----- ----- --- --NOTE - --- ------ --- Only required to be met during a THERMAL MchmA4h' POWER increase or recirculation flow increase in MODES 1 and 2 with one idle recirculation loop when THERMAL POWER is s 30% RTP or when operating loop flow is s 50t rated loop flow. Verify the difference between the bottom Once within head coolant temperature and the RPV steam 15 minutes space coolant temperature is s 145 F. prior to a THERMAL POWER increase or recirculation flow increase l SR 3.4.10.6 - ------- - NOTE- - - ----- --- ---- Only required to be met during a THERMAL POWER increase or recirculation flow (AcTionatE) h increase in MODES 1 and 2 with one non-isolated idle recirculation loop when THERMAL POWER is s 30% RTP or when operating loop flow is s 50% rated loop flow. l Verify the difference between the reactor Once within coolant temperature in the idle 15 minutes ; recirculation loop and the RPV coolant prior to a ! temperature is s 50*F. THERMAL POWER increase or recirculation flow increase FERMI - UNIT 2 Page 3.4-25 (INSERT) . REVISION 10, 07/09/99l 1
i-RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.2 REQUIREMENTS (continued) A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. Perfoming the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be { 1 exceeded between the time of the Surveillance and the time
)
of the control rod withdrawal. h sa 3.4.10.3 w .sR 3.4.10.4 .J R 7,'/,/0 f' a d 5 # I//0 # C_ __- - Differential temperatures within the applicable MtR-limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allow Ir. Pities., :----Hr.:: it;. ihe>= iisiis h.ensur;ances.; thet the-asst.ri, tie... 'vi the anai I,uer :f c.1;i. .. ....i.i;... i.. , g.r. ;; .ysis ..for-;at4sf4ed tne star 6 p 6 M' D dJ hf:=t:; th Ern!!P :: MtN5:astnutus=before k { steriiny i; 14h n:*-reh* %;=pid== =4===te
'Isaurqc,eIna sne gimiin will . ^ ,. . . z f:f,;:' ::- the
=_; .. Ine-tn:c n- = ...._,-
GR 3{/_Op An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 s to compare the temperatures of the operating recirculation loop
-- and.the idle loop.
- QRs hove. -SR 3.(;3' . 3. b: been modified by Nohatrequire Surveillance to be performed on1 in ~ 2, 2, 3 f:-f the?
- DE S T
.Tnsed wi%-ceostere== d- awee e4 ~, p;i . In overali stress on limiting components is Iower. MODE 5, the Therefore, h f8 3 4.tD-V . _
# """"'" "'I"'"I"d" SR 3.4.10 SR 3.4.10 and SR 3.4.10 h Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits (continued) {
BWR/4-STS- B 3.4-53. Rev.4-04107/95--- i 6 Ren0
RCS P/T Limit B 3.4.10 Insert' B 3.4.10 - 3 Limiting differential temperatures within the applicable limits during a THERMAL POWER increase or recirculation flow increase in single loop operation, while THERMAL POWER s 30% k RTP or operating loop flow s 50% of rated loop flow, ensure that resulting thermal stresses will not exceed design allowances. Performing the Surveillance within 15 minutes before starting the idle recirculation pump. THERMAL POWER increase during single loop operation, or recirculation flow increase
- during single loop operation, provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start. power increase, or flow increase.
Insert B 3.4.10 - 4
... for SRs 3.4.10.3 and 3.4.10.4 in MODE 5. In MODES 3, 4 and 5. THERMAL POWER increases are not possible, and recirculation flow increases will not result in additional stresses. Therefore AT limits are only required for SRs 3.4.10.5 and 3.4.10.6 in MODES 1 and 2. The Notes also state that the SR is only required to be met during the event of concern (e.g. pump startup, power increase or flow increase) since this is when the stresses occur.
l l l s FERMI - UNIT 2 Page B 3.4 53 (INSERT) REVISION 10. 07/09/99l
l l JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.4.10 - RCS P/T LIMITS
\
NON BRACKETED PLANT SPECIFIC CH/NGES P.1- These changes are made to NUREG 1433 to reflect Fermi 2 current licensing basis: including design features, existing license requirements and comitments. Additional rewording. reformatting. and revised numbering is made to incorporate these changes consistent with Writer's Guide conventions. Refer to CTS Discussion Of Changes to the related requirement for a detailed justification of changes made to the current licensing basis which are also reflected in the ITS as presented. l P.2 Bases changes are made to reflect plant specific design details, equipment terminology, and analyses. P.3 Not used. P.4 NUREG 1433 Bases do not include discussion of NUREG-1433 3.4.10 Action C Note. Appropriate Bases are included for completeness. P.5 The reference to the NRC Policy Statement has been replaced with a
, more appropriate reference to the Improved Technical Specification
" split" criteria found in 10 CFR 50.36(c)(2)(ii).
P.6 ITS SRs 3.4.10.5 and 3.4.10.6 are added to reflect the specific coolant temperature limits applicable to thermal power and flow increases while in single loop operation. These limits were located in the CTS Specification for recirculation loops operating but are more appropriately located in the Specification for Pressure and Temperature limits. This location for these SRs results in applying the appropriate Actions. This is also the subject of a pending generic change to NUREG-1433. GENERIC CHANGIS C.1 TSTF 35: NRC approved change to NUREG 1433. FERMI UNIT 2 1 REVISION 10. 07/09/99l
l INSERT THIS PAGE IN FRONT OF VOLUME 11 i Volume 11: CTS MARKUP COMPILATION - '
)
Remove Replace 3/44-1(3.4.1 CTS M/U) pg 2 of 6 Rev 2 3/44-1(3.4.1 CTS hW) pg 2 of 6 Rev 10 j 3/44-2(3.4.1 CTS hW) pg 3 of 6 Rev 4 3/4 4-2 (3.4.1 CTS hW) pg 3 of 6 Rev 10 3/4 4-2 (3.4.10 CTS M/U) pg 2 of 8 3/4 4-2 (3.4.10 CTS hW) pg 2 of 8 Rev 10 3/4 4-11 (3.4.4 CTS hW) pg 2 of 2 3/4 4-11 (3.4.4 CTS M/U) pg 2 of 2 Rev 10 I 1 i Res10 07/09/99
S PEco r <eihJ M .! 3 /4. 4 REACTOR COOLANT SY' STEM
'g 3/4.4.1 RECIRCULATION SYSTEM 1 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION Leo 39j ?.'. .:- Two reactor coolant system recirculation loops shall be in operation.
APPLICABILfTY: OPERATIONAL CONDITIONS 1 and 2*. ACTION:
- a. With one reactor coolant system recirculation loop not in operation:
LCo M. l - 1. Within 4 hours: LAil tJOTT r.---. s r)a Pla the individual re culation pump moflow _ o atino ret.f rculatio pump in the Manual L cof roller' for ythe ] g y g'q,g,Qb) Reduce THERMAL POWER to less than or equal to 67.27. of RATED $ THERMAL POWER. j ,j
) Limit the speed [f the operating recircydation pump to lesp/t han) or eaual to 7 5 of rated pump speed e i
d) Inc ase Ine Minin CRITICAL PDWLK Mali (ril,rK) baTeIy L1 L1 th value for sing loop operation reg red by Specifica on
.l.2. s y
e) Change the Average Power Range Monitor (APRM) Simulated Thermal , ' Power - Upscale Flow Biased Scram F.r4 M "M; Mr irire'-ts? O N'O'! WA110wable Values to those applicable for single recirculation ~ ! loop operation per Specifications 2.2.1 and 3.3.6. l f) Perform Surveille.nce Requirement 4.4.1.1.4 if THERMAL POWER is ggg fah less than or equal to 307. of RATED THERMAL POWER or the g,4 ,,, recirculation loop flow in the operating loop is less than or equal to 507. of rated loop flow.
-2. Ot h: .:' -- ha 4-it 1::: "0T ;;;;;;;' . n m nin ine u m ;; ;,_ ,..
- b. With no reactor coolant system recirculation loop in operation while in MdD OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.
- c. With no reactor coolant system recirculation loops in operation, while in h CTlDb) C, OPERATIONAL CONDITION 2, initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours. ,
See etialTe3(Exception 3.f0.4.] I e FERMI - UNIT 2 3/4 4 1 Amendment No. JJ,64,EJ,EJ, EJ,Jpp, 122 PAGE 2. OF 06 Raulo i
e i SPEt tFwAng4 3.% I (Aisw sat paciRutre 3 4o) (Aho sat spcil;ukm s E t ) REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS S ee. T4.1.1;I Each pump discharge valve shall be demonstrated OPGRABLE by
@w 3,g, ,
icycling each valve through at least one complete cycle of full travel at least once per 18 aienths.
/ 4,1
. .;.2 eus s w
.=.A.1.3 With reactor coolant s tem recm.u mien eration,'at I st once per 12 hou erify that: ivobnotin THERMAL POWER i ess than or equal to 6 .2% of RATED THERMAL POWER, nd
- b. The individu recirculatioa pump f1 controller for operating irculation pump is in e Manual mode, a
- c. The speed f the operating recircu tion pump is le or eoual o 75% of rated aa-a une - tha#nl r
j 4.4.1.1.4 With one reactor coolant system loop not in operatjon with THERMAL POWER less than or equal to 305 of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 505 of r$ted loop flow, verify the following differential temperature requirements are met within no more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase:
-- a.
M Less than or equal to 145'F between reactor vessel steam 4 space coolant and bottom head drain line coolant, and i ged{gh b. Less than or equal to 50'F between the reactor coolant within the loop not in operation and the coolant in the 3Ya reactor pressure vessei+*. and
- c. Less than or equal to 50*F between the reactbr coolant within the loop not in operation and the cperating loop.**
l l l N*y. j T .u. - **Reautrement does not apply when the~ recirculation loop not ih operation is
$ip.4 Gu.he- isolated from the reactor pressure vessel.
i 3 M .t6 FERMI - UNIT 2 3/4 4-2 Amendment No. 'U. $9. 57. UB.133 l PE 3 _OF 06 Redl0
S fEttPtewnw 3.4.10 \ (Also see S)ecTGcafim 54.1) ) REACTOR COOLANT SYSTEM [A u s % .9 eci A ca1 % 8 E l) SURVETLtANCE REOUTREMENTS
- 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPGRABLE by Sl dab cycling each valve through at least one complete cycle of full travel at least 351 once per 18 months. '
i.4.1.1.2 DELETED
'4.4.1.1.3 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours verify that: '
g a. THERMAL POWER is less than or equal to 67.2% of RATED THERMAL POWER, and
% ciffa2h % b. The individual recirculation pump flow controller for the 3.Y.I - operating recirculation pump is in the Manual mode, and
- c. The speed of the operating recirculation pump is less than or equal to 75% of rated pump speed.
4.4.1.1.4 With one reactor coolant system loop not in operation with THERMAL POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated loop flow, verify the following differential temperature requirements are met within no j more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase: i i 6A3. 4.lD8 a. Less than or equal to 145'F between reactor vessel steam
, space coolant and bottom head drain line coolant, and
- b. Less than or equal to 50*F between the reactor coolant M 3.4.10 4 witttin the loop not in operation and(the cooiant in th Qeictor pressure vesset", ano N
- c. Less than or equal to 50*F between the reactbr/ coolant l within the loop not in operation and Gne operating loop)**
-* NM und. -
**Recuirement does not apply when the recirculaticn loop not ih operation is 3/f.i0.@ isolated from the reactor pressure vessel.
FERMI - UNIT 2 3/4 4-2 Amendment No. 53. 49. E7 UB.133 PAGE c1 0F 08 REY 0
-e i-OC/ C4 /d/L - -
3 i REACTOR COOLANT SYSTEN LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued)
- e. In OPERATIONAL CONDITION I, with any reactor coolant system I h[ oft 8 UNIDENTIFIED LEAKAGE increase greater than 2 gpm within any 24 hour period, identify the source of leakage increase as not service sensitive Type 304 or 316 austenttic stainless steel within a haneiror be in at least HOT SHUTDOWN within the next 12 Jc6'on C hours and in COLD SHUTDOWN within the following 24 hours.
- f. In OPE 2 and ith any ctor coo nt system % [. , 2 UNID p ,
FIED L ident LCONDJ%ncrease
.i ater tha gpa wit any 4 ur '
3 I he sour f leakage rease at serv nsitive 304 or austeniti ainles eel wi 4 hours in at - t HOT SHUT withi e next in SHUTDOW thin the fo wing 24 ours y hoursand}) SURVEILLANCE RE001REMENTS N *' The reactor coolant system leakage shall be demonstrated to be 3 A.M within each of the above limits by: ! o
& e ne d en W -+ " ~ V vit[6tleastonceperenoursp Mu LA./
bfoMtbrinn,tne ortmary coritetsmaat1umph r hours in vrr.MilDML LUNUlilDN Mnd at least once per 4 hour in OPERATIONAL CONDITIONS 2 and 3 l i di i 3 -
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( FERMI - UNIT 2 3/4 4-II Amendment No. 89 PAGE c2 0F O'2 #dl0
l l l INSERT THIS PAGE IN FRONT OF VOLUME 12 Volume 12: IMPROVED TECHNICAL SPECIFICATIONS . ; I Remove Replace l 3.4.1 ITS pg 3.4-1 Rev 4 3.4.1 ITS pg 3.4-1 Rev 10 3.4.4 ITS pg 3.4-8 Rev 0 3.4.4 ITS pg 3.4-8 Rev 10 3.4.4 ITS pg 3.4-9 Rev 0 3.4.4 ITS pg 3.4-9 Rev 10 3.4.6 ITS pg 3.4-13 Rev 4 3.4.6 ITS pg 3.4-13 Rev 10 l 3.4.10 ITS pg 3.4-26 Rev 0 3.4.10 ITS pg 3.4-26 Rev 10 1 Rev10 07/09/99
g q L . w, Recirculation Loops Operating 3.4.1 l
) 3.4 REACTOR, COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating d i
h LCO 3.4.1 The reactor core shall not exhibit core thermal-hydraulic instability or operate in the " Scram" or " Exit" Regions.
^l 4
8ND
> a. Two recirculation loops with matched recirculation loop W jet pump flows shall be in operation:
k a
- b. One recirculation loop may be in operation provided:
- 1. LC0 3.3.1.1. " Reactor Protection System (RPS)
Instrumentation," Function 2.b (Average Power Range Vl Monitors Simulated Thermal Power-Upscale) Allnwable Value of Table 3.3.1.11 is reset for single loop e operation, when in MODE 1: and k 2. THERMAL POWER is s 67.2% RTP. i
% ............................N0TE - -- - --- -- -- --- -
r Required allowable value modification for single loop Ml operation and THERMAL POWER limitation may be delayed for up L to 4 hours after transition from two recirculation loop operations to single recirculation loop operation. APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation jet pump A.1 Declare recirculation 2 hours loo) flow mismatch not loop with lower flow: ( wit 11n limits. "not in operation." (contint.ed)
-l FERMI UNIT 2 3.4 1 Revision 10 07/09/99 l
t
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c [ l RCS Operational LEAKAGE 3.4.4
) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE:
- b. s 5 gpm unidentified LEAKAGE:
- c. s 25 gpm total LEAKAGE averaged over the previous 24 hour period: and
- d. 5 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1.
APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours not within limit. within limits. Total LEAKAGE not within limit. h B. Unidentified LEAKAGE B.1 Reduce LEAKAGE to 4 hours increase not within t within limits. !
' i mi t-,
d M (continued) ; l FERMI UNIT 2 3.4 8 Revision 10, 07/09/99
.= n-RCS Operational LEAKAGE 3.4.4
') ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME b3 B. (continued) B.2 Verify source of. 4 hours
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unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A ANQ . or B not met. C.2 Be in MODE 4. 36 hours . l DB Pressure boundary LEAKAGE exists. SURVEILLANCE REQUIREMENTS
=
SURVEILLANCE FREQUENCY l jfl SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE 8 hours and unidentified LEAXAGE increase are within limits. ; 1 I i I l FERMI - UNIT 2 3.4 9 Revision 10 07/09/99
F ) I RCS Leakage Detection Instrumentation t 3.4.6 j
] -3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 'RCS Leakage Detection Instrumentation
/'
1 LCO.3.4.6 The following RCS leakage detection instrumentation shall be ' OPERABLE:
- a. Drywell floor drain sump flow monitoring system:
~
Ld b. 8 The!oringsystemchannel:rimary moni and containment atmos here gaseous ra
- c. Drywell floor drain sump level monitoring system.
APPLICABILITY: MODES 1. 2. and 3. i
' ACTIONS-
..................................N0TE -- -- -- --- - -- - --------- --
LC0 3.0.4 is not applicable.
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4 CONDITION REQUIRED ACTION COMPLETION TIME
- A. ;Drywell floor drain A.1 Restore drywell floor 30 days sump flow monitoring drain sump flow
. system inoperable. monitoring system to OPERABLE status.
B. Required primary B.1 Analyze grab samples Once per containment atmosphere of primary 24 hours gaseous radioactivity- containment monitoring. system atmosphere. inoperable. (continued)
. l
,/ I
- l' FERMI -' UNIT 2 3.4 13. Revision 10. 07/09/99
RCS P/T Limits 3.4.10
;- SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY
.SR 3.4.10.5 ----.....--- -- ---NOTE------ ------ -- -
Only required to be met during a THERMAL POWER increase or recirculation flow
/f increase in MODES 1 and 2 with one idle recirculation loop when THERMAL POWER is s 30% RTP or when operating loop flow is Q s 50% rated loop flow.
1l Once within Verify the difference between the bottom 15 minutes l head coolant temperature and the RPV steam prior to a space coolant temperature is s 145'F. THERMAL POWER l increase or recirculation l flow increase SR 3.4.10.6 -- -...--
---. -- NOTE---- -----..---
Only required to t,e met during a THERMAL POWER increase or recirculation flow increase in MODES 1 and 2 with one non-isolated idle recirculation loo) when THERMAL POWER is s 30% RTP or w1en operating loop flow is s 50% rated loop flow.
........................................... Once within 15 minutes l Verify the difference between the reactor prior to a coolant temperature in the iole THERMAL POWER
/ recirculation loop and the RPV coolant increase or 3f temperature is s 50*F. recirculation I flow increase (continued) l FERMI - UNIT 2 3.4 26 Revision 10. 07/09/99
,,e INSERT THIS PAGE IN FRONT OF VOLUME 13 Volume 13: IMPROVED TECHNICAL SPECIFICATIONS BASES .
Remove Replace B 3.4.1 ITS pg B 3.4.1-4 Rev 2 B 3.4.1 ITS pg B 3.4.1-4 Rev 10 B 3.4.3 ITS pg B 3.4.3-2 Rev 0 B 3.4.3 ITS pg B 3.4.3-2 Rev 10 B 3.4.4 ITS pg B 3.4.4-4 Rev 0 B 3.4.4 ITS pg B 3.4.4-4 Rev 10 B 3.4.4 ITS pg B 3.4.4-5 Rev 0 B 3.4.4 ITS pg B 3.4.4 5 Rev 10 B 3.4.10 ITS pg B 3.4.10-7 Rev 0 B 3.4.10 ITS pg B 3.4.10-7 Rev 10 l l l l l l Rev 10 07/09/99 L ..
ir< . Recirculation Loops Operating B 3.4.1
') . BASES APPLICABLE SAFETY ANALYSES (continued)-
Thermal hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal-hydraulic instability are necessary to provide reasonable assurance that the requirements of Reference'4 are satisfied. Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LC0 .Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.2 to ensure that during a LOCA caused by a break of m; the piping of one recirculation loop the assumptions of the i LOCA analysis are satisfied. With the limits specified in kJ SR 3.4.1.2 not met, the recirculation loop with the lower flow must be considered not in o With only one
.% recirculation loop in operation,peration.
modifications to the APRM l 5 Simulated Thermal. Power-Upscale setpoint (LC0 3.3.1.1) and N 6 a limitation on THERMAL POWER may be applied to allow continued operation consistent with the assumptions of the safety analysis. l r Operations that exhibit core thermal hydraulic instability are not permitted. Additionally, in order to avoid G potential power oscillations due to thermal hydraulic instability operation at certain combinations of power and 3 cc
- flow are not permitted. These restricted power and flow regions are referred to as the " Scram" and " Exit" regions y' and are defined by Bases Figure B 3.4.1-1.
A Note is provided to allow 4 hours following the transition to single loop operation from two loop operation to 4 establish the APRM Simulated Thermal Power - Upscale 1l- setpoint in accordance with the single loop allowable value. A which is specified in Table 3.3.1.1-l' and to establish T operation at s 67 2% RTP. The 4 hour period is sufficient
" .to make the sdjustments given the relatively small change-required. - This transition'that results in applying the new b single loop allowable values to APRM OPERABILITY. is such l
1g that any ARPM non compliance with the required allowable
>l value after this 4 hour allowance results in ACTIONS of 4 LCO 3.3.1.1 being entered: no ACTION of LC0 3.4.1 would apply. j
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1 l FERMI' L UNIT. 2 B 3.4.1 - 4 Revision 10. 07/09/99
i n
, e .
i SRVs B 3.4.3
'. BASES APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref.1). For the purpose of the analyses 11 SRVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.
From an over)ressure standpoint, the design basis events are bounded by t1e MSIV closure with flux scram event described above. Reference 2 discusses additional events that are expected to actuate the SRVs. SRVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). 4 LC0 The safety function of 11 SRVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs.1 and 2). The requirements of this LC0 are applicable only to the capability of the SRVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function). t A- l The SRV setpoints, and 3% allowance for setpoint drift, are v established to ensure that the ASME Code limit on peak i reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the UFSAR are also based on these setpoints. Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded. l FERMI - UNIT 2 B 3.4.3-2 Revision 10, 07/09/99
I RCS Operational LEAKAGE B 3.4.4 m. 1 BASES
-APPLICABILITY In MODES 1, 2, and 3 the RCS operational LEAKAGE LC0 a) plies, because the potential for RCPB LEAKAGE is greatest w1en the reactor is pressurized.
In MODES 4 and 5. RCS operational LEAKAGE' limits are not required since the reactor is not pressurized and stresses in the RCPB materials and potential for LEAKAGE are reduced. ACTIONS .Ad With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leak. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours is allowed to reduce the LEAKtSE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and 4 quantified, it may be reclassified and considered as h.- identified LEAKAGE: however, the total LEAKAGE limit would remain unchanged. kl. B.1 and B.2
. An unidentified LEAKAGE increase of > 2 gpm within a 24 hour period is an indication of a potential flaw in the RCPB and must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to be the source of the LEAKAGE increase within the required
.Compietion Time by evaluating service sensitive type 304 and type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased LEAKAGE. This type piping is very susceptible to IGSCC. For an unidentified LEAKAGE increase greater than required limits (in accordance with LC0 3.0.2),
an alternative to this evaluation is to reduce the LEAKAGE increase to within limits (i.e., reducing the LEAKAGE rate such that the current rate is less than the "2 gpm increase in the previous _24 hours" limit: either by isolating the
. source or other possible methods).
The 4 hour Completion Time is reasonable to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety. Ll FERMI UNIT 2- B 3.4.4 -4 Revision 10, 07/09/99 h..
F 1 L l RCS Operational LEAKAGE B 3.4.4 l
] BASES ACTIONS (continued)
C.1 and C.2 If any Required Action and associated Completion Time of , Condition A or B is not met or if pressure boundary LEAKAGE i exists. the plant must be brought to a MODE in which the LC0 ) does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience. to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant safety systems. SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE (e.g., Primary Containment Atmospheric Gaseous Radioactivity. RPV head flange leak detection, and sump monitoring systems). Leakage detection instrumentation is discussed in more l detail in the Bases for LC0 3.4.6. "RCS Leakage Detection ; Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates: however, any . method may be used to quantify LEAKAGE within the guidelines ! of Reference 5. In conjunction with alarms and other 3' i l administrative controls, an 8 hour Frequency for this Surveillance is appropriate for identifying LEAKAGE and for kl tracking required trends (Ref. 6). REFERENCES 1. 10 CFR 50. Appendix A. GDC 30.
- 2. GEAP 5620. April 1968.
- 3. NUREG 76/067. October 1975. ,
- 4. UFSAR. Section 5.2.7.4.3.3.
.5. Regulatory Guide 1.45.
- 6. Generic Letter 88 01. Supplement 1.
j . l FERMI UNIT 2 B 3.4.4- 5 Revision 10 07/09/99
RCS P/i uimits B 3.4.10
) BASES SURVEILLANCE REQUIREMENTS-(continued)
This SR has been modified with a Note that requires this Surveillance to be performed as applicable only during system heatup and cooldown operations and inservice leakage and hydrostatic testing. SR 3.4.10.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. SR 3.4.10.3. SR 3.4.10.4. SR 3.4.10.5. and SR 3.4.10.6 3 Differential temperatures within the applicable limits f ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. Limiting differential temperatures within the applicable limits during a THERMAL POWER increase or recirculation flow increase in single loop operation, while THERMAL POWER s 30% RTP or operating loop flow s 50% of rated loop flow, ensure that thermal stresses resulting from THERMAL POWER increases or recirculation loop flow increases will not exceed design allowances. Performing the Surveillance within 15 minutes before starting the idle recirculation pump. THERMAL POWER increase during single loop operation, or recirculation flow increase during single loop operation, provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start, power increase, or flow increase. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 and SR 3.4.10.6 is to compare the temperatures of the operating recirculation loop and the idle loop. I FERMI - UNIT 2 B 3.4.10 - 7 Revision 10 07/09/99 .}}