ML20211N607
| ML20211N607 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/09/1986 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20211N531 | List: |
| References | |
| 2482K, NUDOCS 8612180263 | |
| Download: ML20211N607 (38) | |
Text
. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _
6 6
ATTAOWENT B TECHNICAL SPECIFICATION CHANGE REQUEST LASALLE COUNTY STATION UNIT 2 PROPOSED CHANGE TO APPENDIX A TECHNICAL SPECIFICATION TO OPERATING LICENSE NPF-18 REVISED PAGES: XIX 3/4 2-5 (Replacement Page) 2-1 3/4 2-6 (Replacement Page)
B 2-1 3/4 3-39 B 2-4 3/4 3-53 B 2-5 3/4 4-1 B 2-6 3/4 4-2 B 2-7 3/4 4-2a (new page) 3/4 2-1 B 3/4 4-1 3/4 2-2 3/4 2-2(a) (new page) 3/4 2-4 2482K 8612180263 PDR 861209 P
ADOCK 05000374 PDR
LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /
CONCENTRATION REQUIREMENTS ........................ 3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na2 Bi oSt s 10 H2O)
VOLUME / CONCENTRATION REQUIREMENTS ................. 3/4 1-22 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE INITIAL(MAPLMGR)
CORE FUELVERSUS AVERAGE TYPES GCRio3, PLANAR GCR233, EXPOSURE, RBl7fo, 9cRBZl9, *D Age tc OCR711 ............................................ 3/4 2-2
=
3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS T AT RATED FLOW .................................. 3/4 2-5 3.2.3-2 K FACTOR .........................................
7 3/4 2-6 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE ....................... 3/4 4-19 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST ........ 3/4 7-33 8 3/4 3-1 REACTOR VESSEL WATER LEVEL ...................'..... B 3/4 3-7 B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1Nev) at 1/4 T AS A FUNCTION OF SERVICE LIFE ..................... B 3/4 4-7 5.1.1-1 EXCLUSION AREA AND SITE B'yNDARY FOR GASEOUS AND LIQUID EFFLUENTS .............................. 5-2 5.1.2-1 LOW POPULATION ZONE ............................... 5-3 6.1-1 CORPORATE MANAGEMENT ..............................
6-11 6.1-2 UNIT ORGANIZATION ................................. 6-12 6.1-3 MINIMUM SHIFT CREW COMPOSITION .................... 6-13 31.1-2 M AttMdM MERAGE PLAPlAA LWCAR. HEAT G Cgggg o y R ATf" (MAPLHGRT VER5uS AvcRAGE ptANAR cxpogggg, Fu r t- TYPE BPS C R B 217 L 3A 2- 2 (M 3.4.1 t - I cece THetmAt PeWER. (7 4 EATeb) veesvs TbfAt.
coRC Food 4% OF RATCQ 3/4 ' 4 - lo.
LA SALLE - UNIT 2 XIX
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS
~
THERMAL POWER. Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam done pressure less than 785 psig or core flow less than 10% of rated flow.
I APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel .
steam done pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4.
THERMAL POWER. High Pressure and High Flow
/
2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not with two recirculation loop cperation and shall not be less than W g th single recirculation loop operation with the reactor vessel steam dose s.ap pressure greater than 785 psig and core flow greater than 10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. -
ACTION: ,
r With MCPR less than +HH' with two recirculation loop operation or less than r
4-9P with single recirculation loop operation and the reactor vessel steas
- F # ose d pressure greater than 785 psig and core flow greater than 10% of rated j flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the require- _
sents of Specification 6.4.
l REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABILIT_Y: OPERATIONAL CONDITIONS 1, 2, 3, and 4.
ACTION:
With the reactor coolant system pressure, as reasured in the reactor vassel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4.
LA SALLE - UNIT 2 2-1 :
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4 2.1 SAFETY LIMITS 8ASES The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the j environs. Safety Limits are established to protect the integrity of these i 1
barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated '
to occur if the limit is r.ot violated. Because fuel damage is not directly M observable, a step-back approach is used to establish a Safety Limit such tha .
/.07 ~ the MCPR is not less than 4r46- for two recirculation loop operation and or single recirculation loop operation. MCPR greater than Qfor two recircula- ;
tion loop operation andJr#-for single recirculation loop operation representsg 1
{ d conservative margin relative to the conditions required to maintain fuel !
#8 cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or l j cracking. Although some corrosion or use related cracking may occur during
- the life of the cladding, fission product migration from this source is incre-mentally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.
i While fission product migration from cladding perforation is just as measurable i
, as that from use related cracking, the thermally caused cladding perforations li (x signal a threshold beyond which still greater thermal strest.es may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, ( PR of 1.0. These conditions represent a signif- ,
icant departure from the conhR. ion intended by design for planned operation. l 2.1.1 THERMAL POWER. Low Pressure o.' Low Flow The use of the GEXL correlation is not valid for all critical power _
calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by i
other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region :
1 is essentially all elevation head, the core pressure drop at low power and '
flows will always be greater than 4.5 psi. Analyses show that with a bundle 4
flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving
- head will be greater than 28 x 108 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors,
- this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
l Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
(
LA SALLE - UNIT 2 8 2-1
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BASES TABLE 82.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT
- STANDARD .
DEVIATION QUANTITY (% of Point)
Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow Two Recirculation Loop Operation 2.5
- Single Recirculation Loop Operation 6.0 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings Two Recirculation Loop Operation .JM4P" E 7 l Single Recirculation Loop Operation 6.8 R Factor M /. 6 Critical Power 3.6 "The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.
The values herein apply to both two recirculation loop operation and single recirculation loop operation, except as noted.
LA SALLE - UNIT 2 8 2-4 ,
, - , - -v,,,----~- - - - - - c,,n ,,-,,-,---.-,e-,-.-- , - - - - - - - - - - - - . - - - - , -,,- - , , - - - - , - -
4 Bases Table 82.1.2-2 .
NOMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLADOING INTEGRITY SAFETY LIMIT THERMAL POWER 40 ; ;; 3293 MW Core Flow 4 00. ; n; /... 10 1 5 MIL / Ac
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Dome Pressure 1010.4 psig
. .... . 2
,<.....,.,,..__m. ...
R-Factor -ue;6 ;; 7 5 ;;,,,,,,; ;,;;;
e... i c.x t - 1. 00^
":di x L= :- if;;nt '.";0
- f. css -
Q GWl/t
/. 03/ -
7 GWd/t t.030 -
15 sca/g
/.033 20 GWd/t C
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Bases Table 82.1.2-3 RELATIVE BUNOLE POWER DISTRIBUTION USED IN THE GETAB STATISTICAL ANALYSIS Percent of Fuel Bundles Within Range of Relative Bundle Power Power Interval _
i:iii:i:in 4
!?
1.375 to 1.425 O.1 r.i 1.325 to 1.375 2-73 1.275 to 1.325 -t-t- 7. s 1.225 to 1.275 '2 1.s 1.175 to 1.225 -ih4- 7 3 1.125 to 1.175 --6:4- ti.:
1.075 to 1.125 -+.4- 4 7 1.025 to 1.075 6-47
<1.025 M 41.C
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LA SALLE - UNIT 2 B 2-6
Bases Table B2.1.2-4 R-FACTOR DISTRIBUTION USED IN GETAB STATISTICAL ANALYSIS 8x8 Rod Array R-Factor
<-4Hg$r - 2 :i .e. -ter C r -i c r ;; r.t Cr-icb.;;.% !- ::b rt Rod Sequence No.
..-v.
-,. ..-. I
-- ,...m,. , . , ...
.v,a -
-..... o.038 ..... 2 1.eei -1. 000 /.e37 ' ' ' ' ^
. 3
-1.042 -l.000 /CJ7 4 390 - 4
-h406- -hee? /o35 - 1. 0^^, 5
-- 5996- +:Ott t035 ' " ^C-
. 6 4.;;G -hef6- ^c30 + 929- 7
- ;.^^^ -11.025 6 /.oso 2 1^. ,20 8 through 64 O
e LA SALLE - UNIT 2 8 2-7 e .,- - - . -_ - - - - . _ , - , . . - , - . - . - - - - . , , - - - - . - - . - , ,, , , . - . - - -,n-,_.,n._
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION
- and 3.2.1-1 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1. The limits of Figure 3.2.1-1"sha11 be reduced to a value of 0.85 times the two recirculation loop operation limit when in single recirculation loop' operation.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
and 3.2.1 - 2.
WithanAPLHGRexceedingthelimitsofFigure3.2.1-1(initiatecorrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits -
determined from Figure 3.2.1-4/ snA 3.2.1 - Z .
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 'after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.
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=25' 000 ,' ' ' ' 50' 000 AVERAGE PLANAR EXP09bHE Wo/4n) g MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGRI VERSUS AVERAGE PLANAR EXPOSURE INITIAL CORE FUEL TYPES 8176, 62W, Aub R&71 FIGURE 3.2.11 i
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g MAPLHGR VS. Average Planar Exposure '
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! 0 5000 10000 15000 20000 25000 30000 35000 l Average Planar Exposure (mwd /St) t%XIMJM AVERAGE PLANAR LIEAR EAT GEERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, FUEL TYPE BP8CRB299L FIGURE 3.2.1-2 i
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I POWER DISTRI'BUTION LIMITS * . 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICA( POWER RATIO (MCPR) shall be squal to or greater than the MCPR limit determined from Figure 3.2.3-1 times the K, determined from Figure 3.2.3-2 for two recirculation loop operation and shall be equal to or greater than the MCfR limit determined from Figure 3.2.3-1 + 0.01 times the K determined from Figure 3.2.3-2.for single recirculation loop operation, previg _th:t th: : d ef :y:1: r;;ir;;ictier pu=p trip-{E0E,-RPE :y:^-- is - __ ,___ m __ __,,,, l n o e.n . . . e.............o.....
.. ....m...
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION
- e. With the end ef ;ycle recirculatica pu:p trip ;y:te: 'ne? r:ble per Sp;;ificatica 0.0.4.0, eperati;r. ::y ;;ntin : d the pre"iffen: ef Specificatien 0.0.4 ere net applic; tie preeid:d th:t, with*" 1 h r.
TC^ i . d.;.. . M.o ;u L. .yu.'. ;e e. s.eeter th.. the 5"" 'i;it
- h=- '- Ff;;r: 2.2.3-1 50C ..T in:;:r:b?: cere:, tir:: the y the--
in Figur; 3.2.1-2.
-A With MCPR less than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours or reduce THERMAL POWER to less than 25% of. RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with: l a. t*** = 0.86 prior to performance of the initial scram time measurements l for the cycle in accordance with Specification 4.1.3.2, or
- b. t,y, determined within 72 hours of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2:
- a. ,
At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL R00 PATTERN for MCPR.
LA SALLE - UNIT 2 3/4 2-4 l
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INSTRdMENTATION - ENO-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION l l LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system 'instrumenta-tion channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE
; TIME as shown in Table 3.3.4.2-3. ,
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of PATED THERMAL POWER. ACTION:
- a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable'until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels one less than r.equired by the Minimum OPERABLE Channels per Trip System requirement for one or
~
both trip systems, place the inoperable channel (s) in the tripped j condition within 1 hour. .
- c. With the number of OPERABLE channels two or more less than required i by the Minimum OPERABLE Channels per Trip System requirement (s) for one trip system and:
- 1. If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 1 hour.
! 2. If the inop'erable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable. i
- d. With one trip system inoperable, restore the inoperable trip system 1
t to
~ OPERABLE
>n-- m . ,,, status within 72 hours or rehcc t-ke the ACTIO"h r;56Ruir-d T HCem P0 foby le6S $ $* A Me]'k' dis litSMt. Twca Orlta the n ed 6 houG.
- e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour or tek- the ACTIO" required by SpecifiG6tish 3.2.0 .
te "Le TifGR ML PaMEf to leu N#^
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LA SALLE - UNIT 2 3/4 3-39
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.- 1, 3/4.4 REACTOR COOLANT SYSTEM l 1
3/4.4.1 RECIRCULATION SYSTEM , 1 RECIRCULATION LOOPS 1 LIMITING CONDITION FOR OPERATION s N, i 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation. ' , , APPLICABILITY: OPERATIONAL CONDITIONS 18 and 2". ACTION:
- a. With one reactor coolant system tracirculation loop not in operation:
~
- 1. Within 4 hours:
u a) . Place the recirculation flow control system in the Master "
. Manual mode, andh b) hdce "i!""." L " obc" t; ; 53 ;f "'T50 "C"."." L ^C"J ", ;nd ,
O -e). Increase the MINIMUM CRITICAL POWER RATIO (MCPR') Safety Limit by 0.01 to i- W per Specification 2.1.2, and, s.o t c') -d)- Increase the MINIMUM CRITICAL POWER AATIO (MCPR) Limiting Condition for Operation by 0,pfger specification 3.2.3, and, M -e). ReducetheMAXIMUMAVERAGEPLkNARLINEARHEATGENERATIONRATE (MAPLHGR) Itait to a value of 0.85 times the two recirculation loop operation limit per Specification 3.2.1, and, _
\
e) -4). Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single loop recirculation loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6. A^. le;Z en;; p;7 12 M .70: , a) Ver t the APRM flux noise averaged over nutes does not exc esk to peak; oth, , reduce the 48c44^J recirculation loop ti RM flux noise is less
/>af e d, than the 5% peak to is ,
b) Ver at the core plate AP noise does no d 1 psi eak to peak; otherwise, reduce the recirculation lo fk .atil the # acise is less th;n th;.1 ;;;f limit.
*See Special Test Exception 3.10.4.
LA SALLE - UNIT 2 3/4 4-1
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Following pag'c N/4 4-1:
- i
- 2. When operating within the surveillance.'egion r specifisc in Ftgure 3.4'.1.1-1: '
- a. With core fpw less than 39% of rated core
, - ficw, init' idee action within 15 minutes to
-T either: '
s"
') ' 'm 1. Leave the surveillance. region within 4 hours, or '
Ls ( i
,\ 2. Increase core,fldw *o greater than or 4
, equal to 39% of' rated flow within 4 4
- bours.
s ;, J . b.'With the ApRM and LpRM# neutron flux noise level greater than three (3) t ir.ies their Q established baseline noise levels:
'\
i +
- 1. Initiate corrective action within 15
~ minutes to restore the noise levels to P within the required limit within 2 3 hours, otherwise 4
3
,' 2. Leavegthe surveillance region specified s,, in Figure 3.4.1.1-1 within i.i.: next 2 hours.
\,
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s
# - Detector levels A and C of one LpRM string per core octant plus detector levels A and C of one LpRM string in the center region of the core should be monitored.
4 4 I i
REACTOR COOLANT SYSTE LIMITING CONDITION FOR OPERATION (Continued) l ACTION: (Continued)
- 3. The provisions of Specification 3.0.4 are not applicable.
- 4. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- b. With no reactor coolant system recirculation loops in operation, immediately initiate seasures to place the unit in at least HOT SHUTDOWN within the next 6 hours. -
SURVEILLANCE REQUIREMENTS
\
4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at lent once per 18 months by:
- a. Verifying that the control valve fails "as is" on loss of hydraulic pressure at the hydraulic power unit, and
- b. Verifying that the average rate of control valve movement is:
. 1. Less than or equal to 11% of stroke per second opening, and
- 2. Less than or equal to 11% of stroke per second closing.
'$erT n' /
follod!'] Pa]' LA SALLE - UNIT 2 3/4 4-2
Following page 3/4 4-2 4.4.1.2 With one reactor coolant system recirculation loop not in operations
- a. Establish baseline APRM and LPRM# neutron flux noise level values within 4 hours upon entering the surveillance region of Figure 3.4.1.1-1 provided that the baseline values have not been established since last refueling.
- b. When operating in the surveillance region of Figure 3.4.1.1-1, verify that the APRM and LPRM*
neutron flux noise levels are less than or equal to three (3) times the baseline values:
- 1. At least once per 12 hours, and
- 2. Within 1 hour after completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER, initiating the surveillance within 15 minutes of completion of the increase.
- c. When operating in the surveillance region of Figure 3.4.1.1-1, verify that core flow is greater than or equal to 39% of rated core .
flow at least once per 12 hours.
# - Detector levels A and C of one LPRM string per core octant plus detector levels A and C of one LPRM string in the cente.- region of the core should be monitored.
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3/4.4 REACTOR.CGOLANT SYSTEM , BASES i 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and been found to be acceptable i.-'.; t.t nt '=1 y ': : 'r, provided the unit is operated in accordance with the single recirculation loop operation Technical Specifications herein. An inoperable jet pump is not, in itself, a suffici'eht reason to declare i a recirculation loop inoperable, but it does present a' hazard in case of a design-basis-accident by increasing the blowdown area and reducing the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed scheduled for significant degradation. Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criterion. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. Where the recir-culation loop flow mismatch limits can not be maintained during the recir-culation loop operation, continued operation is permitted in the single recirculation loop operation mode. In order to prevent undue stress on the vessel nozzles and bottom head . - - region, the recirculation loop temperatures'shall be within 50*F of each other prict to startup of an idle. loop.- The loop temperature must also be within. .. 50*F of the reactor pressure vessel coolant temperature to prevent thermal . shock to the recirculation pump and recirculation nozzles. Since the coolant < in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145'F. Lser6 folloa V3
~~
3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 18 OPERABLE safety / i relief valves is required to limit reactor pressure to within ASME III l allowable values for the worst case upset transient. - Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of j Specification 4.0.5. I
*.?
LA SALLE - UNIT 2 B 3/4 4-1
Insert on page B 3/4 4-1 The possibility of thermal hydraulic instability in a BWR has been investigated since the startup of early BWRs. Based on tests and analytical models, it has been identified that the high power-low flow corner of the power-to-flow map is the region of least stability margin. This region may be encountered during startups, shutdowns, sequence exchanges, and as a result of a recirculation pump (s) trip event. To ensure stability, single loop operation is limited in a designated r?stricted region (Figure 3.4.1.1-1) of the power-to-flow map. Single loop operation with a designated surveillance region (Figure 3.4.1.1-1) of the power-to-flow map requires monitoring of ApRM and LPRM noise levels. O i 1
I I TABLE 3.3.6-2
; 9 CONTROL R00 WITH0RAWAL BLOCK INSTRUMENTATION SETPOINTS m
r"- E TRIP FUNCTION TRIP SETPOINT ALLOWA8tE VALUE
- 1. R00 BLOCK MONITOR
[
- a. Upscale 3 1) Two Recirculation Loop Operation 10.66 W + 40E 35 % 10.66 W + 4M 417
- 2) Single Recirculation Loop Operation < 0.66W + e 32.7 7. < 0.66W + 37-N- 35 7 70
- b. Inoperative N.A. R.A.
- c. Downscale > 5% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER l
- 2. APRM
- a. Flow Blased $1mulated Thermal Power-Upscale i 1) Two Recirculation R Loop Operation 1 0.66 W + 42%" < 0.66 W + 45%"
- Single Recirculation 2)
T Loop Operation < 0.66W + 36.7%* < 0.66W + 39.7%* 5 b. Inoperative N.A. H.A.
- c. Downscale > 5% of RATED THERMAL POWER > 3% of RATED THERMAL POWER
- d. Neutron Flux-High 317%ofRATEDTHERMALPOWER 514%ofRATEDTHERMALPOWER
- 3. SOURCE RANGE MONITORS
- a. Detector not full in N.A. N.A. 5 5 < 5 x 10 cp,
! b. Upscale < 2 x 10 cps
- c. Inoperative R.A. R.A.
- d. Downscale 1 0.7 cps > 0.5 cps
- 4. INTERMEDIATE RANGE MONITORS
- a. Detector not full in M.A. N.A.
- b. Upscale < 108/125 of full scale < 110/125 of full scale
- c. Inoperative N.A.
R.A.
- d. Downscale > 5/125 of full scale > 3/125 of f'ull scale
*The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.
i
O ATTACHMENT C TECHNICAL SPECIFICATION CHANGE REQUES? LASALLE COUNTY STATION UNIT 2 SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment to Operating License NPF-18 for LaSalle County Station Unit 2 and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.90, operation of LaSalle County Station Unit 2 in accordance with the proposed amendment will not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated because:
For Cycle 2, the MCPR fuel cladding integrity safety limit was changed from 1.06 to 1.07 for two recirculation loop operation, and from 1.07 to 1.08 for single recirculation loop operation. The safety limit is smaller for initial cores because the uncertainties in TIP symmetry and the R Factor are smaller. The addition of a new MAPLHGR limit vs Exposure curve for the reload fuel type BP8CRB299L in the MAPLHGR Specification. The MPALHGR limits were provided by General Electric in the General Electric in the Supplemental Reload Licensing Submittal for L2C2. The replacement of the existing MCPR curve with a revised cure which reflects the limiting transients for Cycle 2. The MCPR limits were provided by General Electric in the Supplemental Reload Licensing i Submittal for L2C2. Although the cycle specific anlaysis demonstrated adequate stability, single loop thermal hydraulic stability monitoring requires were added to the Recirculation System Technical Specification to address NRC concerns in this area.
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated because:
The proposed Technical Specification changes do not represent significant changes in acceptance criteria or safety margins and all changes have been made based on methods that have been previously accepted by the NRC. The reload core involves a new fuel type which has been evaluated in the General Electric Standard Application for . Reactor Fuel (GESTAR). The new fuel type presents no unreviewed safety questions because the bundle design has been evaluated in GESTAR with
- approved methods.
1 I i
O
- 3. Involve a significant reduction in the margin of safety because:
The deletion of the EOC-RPT inoperable provision in the MCPR and EOC Recirculation Pump Trip System Technical Specifications. The BOC-RPT inoperable analysis was not justified for in the second cycle but may be included in future cycles. The replacement of the existing Kg curve with a revised curve which is based on LaSalle's rated core power and core flow. The original curve was a generic curve. Based on the preceeding discussion, it is concluded that the proposed system change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed changes do not constitute a significant hazards consideration. 2482K
e ATTENWrt.C TECHNICAL SPECIFICATION CHANGE REOUEST LASALLE COdNTY STATION UNIT 2 SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment to Operating License NPF-18 for LaSalle County Station Unit 2 and determined that it does not represent a significant
'azards consideration. Based on the criteria for defining a significant
- .a:%rds consideration established in 10 CPR 50.90, operation of LaSalle County Station Unit 2 in accordance with the proposed amendment will not:
- 1. Involve a significant increase in the probability or consec3:ences of an accident previously evaluated because:
For Cycle 2, the MCPR fuel cladding integrity safety limit was changed from 1.06 to 1.07 for two recirculation loop operation, and from 1.07 to 1.08 for single recirculation loop operation. The safety limit is smaller for initial cores because the uncertainties in TIP symmetry and the R Factor are smaller. The addition of a new MAPLHGR limit vs Exposure curve for the reload fuel type BP8CRB299L in the MAPLHGR Specification. The MPALHGR limits were provided by General Electric in the General Electric in the Supplemental Reload Licensing Submittal for L2C2. The replacement of the existing MCPR curve with a revised cure which reflects the limiting transients for Cycle 2. The MCPR lLmits were provided by General Electric in the Supplemental Reload Licensing Submittal for L2C2. Although the cycle specific anlaysis demonstrated adequate stability, single loop thermal hydraulic stability monitoring requires were added to the Recirculation System Technical Specification to address NRC i concerns in this area.
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated because:
i The proposed Technical Specification changes do not represent significant changes in acceptance criteria or safety margins and all changes have been made based on methods that have been previously
- accepted by the NRC. The reload core involves a new fuel type which l has been evaluated in the General Electric Standard Application for Reactor Fuel (GESTAR). The new fuel type presents no unreviewed safety questions because the bundle design has been evaluated in GESTAR with approved methods.
i [
, _ - - ~ _ _ . . _ . _ _ _ _ _ _ _ . _ , _ . , _ _ _ . . . . . _ . . _ _ .__, _ . _ _ . . . _ . _ _ _ . _ _ _ _ _ _ .
e
- 3. Involve a significant reduction in the margin of safety because:
The deletion of the EOC-RPT inoperable provision in the MCPR and EOC Recirculation Pump Trip System Technical Specifica*. ions. The BOC-RPT inoperable analysis was not justified for in the second cycle but may be included in future cycles. The replacement of the existing Kg curve with a revised curve which is based on LaSalle's rated core power and core flow. The original curve was a generic curve. Based on the preceeding discussion, it is concluded that the proposed system change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed changes do not constitute a significant hazards consideration. 2482K
e e ATTACIGENT D TECHNICAL SPECIFICATION CHANGE REOUtIST LASALLE COUNTY STATION UNIT 2 GE SUPPLEMENTAL RELOAD LICENSING SUBMITTAL CYCLE 2 NOTE: 7 pages of additional information is attached, i 2482K
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